ML20217E938
| ML20217E938 | |
| 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-014, A-PENG-CALC-014-R00, A-PENG-CALC-14, A-PENG-CALC-14-R, NUDOCS 9710070315 | |
| Download: ML20217E938 (109) | |
Text
Arkansas Nuclear One - Unit 2 Pilot Plant Study Risk-Informed Inservice Inspection Evaluation for 1:he Chemical and Volume Control System September 1997
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A PENG CALC-014 Rzvlsl:n 00 Design Analysis Title Pcge O
Title:
Implementation of the EPRI Risk Informed Inservice Inspection Evaluation D
Procedure for the Chemical Volume Control System at ANO 2 Document Number:
A-PENG-CALC-014 Revision 00 Number:
Quality C' ass:
O QC 1(Safety Related)
O QC 2 (Not Safety-Related) @ QC 3 (Not Safety-Related)
- 1. Approvalof Completed Analysis This Design Analysis is complete and verified. Management authorizes the use ofits results.
Printed Name Signature, Date j
j Sh Cognizant Engineer (s)
R. A.Weston A.V.Bauer g [q Mentor g None
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Independer t Reviewer (s)
R. E. Jaquith 2
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Management Approval B. T. Lubin Project Manager
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- 2. Package Contents (this section may be completed after Management approval):
Total page count, including body, appendices, attachments, etc.
108 List associated CD-ROM disk Volume Numbers and path names:
3 None Note: CD ROM are stored as separate Quality Records CD-ROM Volume Path Names (to lowest directory which uniquely applies to this document)
Numbers Total number of sheets of microfiche:
@ None Number of sheets:
Other attachments (specify):
- 3. Distribution:
B. Boya (2 copies) i V
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ll A PENG CALC 014 Rsvision 00 Psgs 2 of S3 RECORD OF REVISIONS Rev Date Pages Changed Prepared By Approved By 00
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Calculation No. A.PENG. CALC.014, Rev. 00 Page 3 of 53 IABLE OF CONTENTS SECTION PAGE
- 1. 0 PURPOSE...............................................................................................................5 2.0 SCOPE...................................................................................................................S 3.0 SYSTEM IDENTIFICA TION AND BOUNDARY DEFINITION............................................ 6 3.1 S YS TEM DESCRIP TION................................................................................... 6 3.2 S YS TEM B O UNDA R Y...................................................................................... 7
- 4. 0 CONSEO UENCE EVA L UA TlON............................................................................... 1 7 4.1 CONSEQ UENCE A SS'l*a1P TIONS..................................................................... 19 4.2 CONSEQ UENCE IDENTIFICA TION................................................................... 21 4.3 SHUTDO WN OPERA TION AND EXTERNAL EVENTS........................................... 21 5.0 DEGRA DA TION MECHA NISMS EVA L UA TION........................................................... 28 5.1 DA MA GE GR O UPS......................................................................................... 2 9 5.2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICA TION........................... 29 S.3 BASICDATA................................................................................................41
- 6. 0 SER VICE HIS TOR Y A ND SUSCEP TIBILITY REVIEW.................................................... 42
- 7. 0 RISK E VA L UA TION................................................................................................ 4 5
- 8. 0 EL EMEN T SEL EC TION............................................................................................ 4 8
9.0 REFERENCES
.........................................................................................................S2 p)
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LIST OF TAULES NUMBER PAGE 1
C VCS B O UNDA RIES.............................................................................................. 13 2
C VCS CONSEQ UENCE A SSESSMENT SUMMA R Y..................................................... 24 3
CVCS CONSEQUENCES, FIGURES AND ISOMETRIC DRA WINGS................................. 25 4
DA MA GE GR O UPS................................................................................................ 2 8 5
DEGRADA TION MECHANISM CRITERIA AND SUSCEPTIBLE REGIONS......................... 30 6
SERVICE HISTORY AND SUSCEPTIBILITY REVIEW - CHEMICAL AND VOLUME CONTROL SYSTEM...............................................................................................................44 7
RISK SEGMENT IDENTIFICA TION........................................................................... 4 6 8
RISK INSPEC TlON SC0PE....................................................................................... 48 9
ELEMENT SELEC TION RISK CA TEGOR Y 4.............................................................. 49 10 ELEMENT SELEC TION - RISK CA TEGOR Y 2.............................................................. 51 LIST OF FIGURES NUMBER PAGE 1A CHEMICAL AND VOLUME CONTROL SYSTEM (CHARGING PORTION)......................... 14 IB CHEMICAL AND VOLUME CONTROL SYSTEM (BORIC AC/D MAKEUP PORTION)..........15 IC CHEMICAL AND VOLUME CONTROL SYSTEM (LETDOWN PORTION).......................... 16
/N 2
SCHEMA TIC OF CVCS CHARGING FLOW PA TH FROM REGEN. Hx.............................. 26
- -*)
3 SCHEMATIC OF CVCS LETDOWN FLOW PATH TO LETDOWN CONTROL VALVES........ 27 ABB Cornbustion Engineering Nuclear Operations
MIDIF l
Calculation No. A.PENG-CALC-014, Rev. 00 Page 4 of S3 LIST OF APPENDICES A
FMECA - CONSEQUENCE INFORMA TION REPORT B
FMECA - DEGRADA TION MECHANISMS C
FMECA SEGMENT RISK RANKING REPORT D
QUAUlY ASSURANCE VERIFICA TION FORMS O
O ABB Combustion Engineering Nuclear Operations
A It R M IDIF Calculation No. A PENG-CALC 014, Rev. 00 L
Page 5 of 53
- 1. 0 PURPOSE The purpose of this evaluation is to document the implementation of the Electric Power Research Institute (EPRI) Risk-Informed Inservhe Inspection Evaluation Procedure (RISI) of Reference 9.1 for the Chemical and Volume Control System (CVCS) at Arkansas Nuclear One, Unit 2 (ANO 2), Entergy Operations, Inc. The RISI evaluation process provides an alternative to the requirements in ASME Section XI for selecting inspection locations. The purpose of RISIis to identify risk significant pipe segments, define the locations that are to be inspected within these segments, and identify appropriate inspection methods.
This evaluation is performed using the guidelines of the EPRI Risk-informed Inservice Inspection Evaluation Procedure of Reference 9.1 and in accordance with the requirements of the ABB Combustion Engineering Naclear Operations Quality Procedures Manual (QPM-101).
- 2. 0 SCOPE This evaluation applies to the CVCS at ANO-2, and utilizes the ISIS Software (Reference 9.2), which has been specifically developed to support and document the implementation of the EPRI RISIprocedure.
As part of the RISI procedure, the system boundaries and functions are identified. A risk evaluation is performed by dividing the system into piping segments which are determined O
to have the same failure consequences and degradation mechanisms.
The failure consequences and degradation mechanisms are evaluated by assigning each segment to the appropriate risk category and identifying the risk-significant segments.
Finally, the inspection locations are selected.
The guidelines used in determining the degradation mechanisms, the failure consequences and the risk significant segments are those described in Reference 9.1.
/V.
ABB Combustion Engineering Nuclear Operations
MIF19 O
Calculation No. A PENG-CALC 014, Rev. 00 Page 6 of 53
- 3. 0 SYSTEM IDENTIFICA TION AND ROUNDARY del'INITION
3.1 System Description
The Chemical and Volume Control System (CVCS) is designed to perform the foHowing functions:
Maintain the chemistry and purity of the reactor coolant during normal operation and during shutdowns, including crud burst cleanup; Maintain the required volume of water in the RCS by compensating for coolant contraction or expansion resulting from changes in the reactor coolant temperature; Provide a controlled path for discharging fluid no the BMS; Control the boron concentration in the RCS in order to maintain acceptable CEA configuration, compensate for reactivity changes associated with bumup and major changes in coolant temperature and xenon concentration, and provide the required shutdown margin to maintain the reactor subcritical; Provide auxiliary pressurizer spray for operator control of pressure during the final stages of shutdown and to aHow pressurizer cooling; Provide a means for functionally testing the check valves which isolate the SIS from the RCS; Provide continuous measurement of reactor coolant boron concentration and fission product activity; 0
CoHect reactor coolant pump seal controHJd bleed-off; O
Leak test the RCS; Inject concentrated boric into the RCS upon a Safety injection Actuation Signa l(SIAS)
Automatically divert the letdown flow to the BMS when the volume control tank is at the highest permissible level; and Provide an alternate means for fining the RCS During normal operations, coolant flow from the cold leg in loop 2P-32A of the RCS passes through the tube side of the regenerative heat exchanger for an initial temperature reduction.
The cooled fluidis reduced to the operating pressure of the letdown heat exchanger by one of two letdown control valves.
The final reduction to the operating temperature and pressure of the purification system is made by the letdown heat exchanger and one of two letdown back pressure valves. The flow then passes through a filter, one of two purification ion exchangers and a strainer, andis then sprayedinto the volume control tank (VCT).
The charging pumps take suction from the VCT and pump the coolant into the RCS. One charging pump is normaHy in operation and one letdown control valve is controHed to maintain the exact balance between letdown flow rate plus reactor coolant pump bleed-off flow rate and charging flow rate. The charging flow passes through the sheH side of the regenerative heat exchanger for recovery of heat from the letdown flow before being
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retumed to the RCS.
Coolant is retumed as chargi... fluid by using ae of the three charging pumps which take suction from the VCT and retum coolant through the regenerative heat exchanger to RCS loops 2P-32B and 2P-32C.
U ABB Combustion Engineering Nuclear Operations l
A Ik It A% FIF Calculation No. A-PENG-CALC-014, Rev. 00 Page 7 of S3 3.2 System Boundary The CVCS is defined consistent with the FSAR (Reference 9.3). The scope of this analysis includes au Class 1 and 2 piping in this system which is currently examined in tioe ANO 2, ASME Section XIInservice Insoection (ISI) Program (Reference 9.6). In addition, the Class 2 piping which is in the CVCS flow path is also included in this evaluation. In identifying the scope for the CVCS evaluation, the lines which are part of or interface with the CVCS were examined to determine their risk significance. The system boundaries are defined in Table 1 and Figures 1(A, B, C).
Certain line segments contain welds which were not entered in the database (Reference 9.2) as outlined below. The majority of these lines are located in Tank and Pump rooms and Corridors flood zone (i.e., RAB-2040-JJ) of the Reactor Auxiliary Building. Based on observations during the walkdown, the ionpact of spatial effects is considered to be insignificant (see Section 4.0).
3.2.1 Letdown drain line downstream of drain volve 2RC SA (2HCC-12")
This line segment provides a drain path from RCS cold leg 2P-32A to the Reactor Drain Tank (RDT). This line is normany isolated from RCS cold leg 2P-32A by the manual drain valves. If the segment remains isolated, a failure in this line segment would not cause an initiating event, and it is not used to accompli?h or support any of the safety functions foHowing a design basis event. If passive failure of both valves were to occur, the segment would be exposed to normal operating conditions. By exposing the segment to normal operating conditions, the potential for a smaH LOCA wists. Because both valves must fail, the resulting consequence of a potential sman LOCA would therefore be LOW. No degradation mechanisms were identified for the welds in this line segment. Because of the LOW consequence and no damage potential, the risk significance of the segment failure would be LOW (i. e.,
CA T 7).
Since no element selections are needed for low risk-significant segments, the welds for this line were not entered in the database.
3.2.2 Letdown line from downstream of letdown control valves 2CV 4816 and 2CV-4817 to the Volume Control Tank (2FCB-212', 2FCB 22 2", 2HCB-45-3', 2HCB-43-3',
2HCB-44-3', 2HCB-66-2", 2HCB SO-2", 2HCB-47-3', 2HCB-49-3", 2HCB 172 3',
2HCB-156 3")
These line segments divert a smaH fraction of the reactor coolant to the Chemical and Volume Control System (CVCS) to provide filtering and purification of the reactor coolant, and to maintain the pressurizer level within its desired range. The diversion or letdown is an ongoing process during normalpower operation. In the event that letdown flow is unavailable, the charging pumps can be operated in a manner that would compensate for RCP bleed-off and would provide makeup as necessary. However, a lack of RCS chemistry control for an extended period would cause a controHed plant shutdown. A failure in any of the above line segments would initiaHy result in excessive letdown flow and a resulting low pressurizer level.
The pressurizer level control system would compensate for the low level by closing the letdown control valve and starting the standby charging pumpfs).
The pressurizer pressure control system would start the backup heaters to maintain pressurizer pressure. There is a high probability that the failed segment would be detected and isolated based on high/ low letdown flow (depending on the break location), unexpected decrease in pressurizer level, low VCT level and excessive ABB Combustion Engineering Nuclear Operations
N A"EIf 59
(
Calculation No. A PENG-CALC-014, Rev. 00 Page 8 of 53 charging. Such conditions are indicated or alarmed in the control room.
The operators can isolate the break from the control room by closing the letdown stop valve 2CV-4820-2 or the regenerative heat exchanger inlet valve 2CV48211.
Because letdown is not needed to accomplish any of the safety functions following plant shutdown and the multiple isolation capabilities, a LOW consequence category would therefore be assigned to a segment failure. Note that during a limiting fault event (i.e., LOCA or steam line break) letdown flow is automaticaHy isolated. No degradation mechanisms were identified for the above segments. Based on the LOW consequence category and no leak potential, the risk significance of a segment failure would be LOW (i.e., CAT 7). Since no element selections are needed for low risk-significant segments, the welds for these lines were not enteredin the database.
3.2.3 Letdown flow path to the Baron Management System (BMS) from downstream of VCTinlet valve 2CV-4826 to upstream of valve 2CVC 3412HCB 52-3')
This line segment provides a flow path for letdown to the vacuum degassifier in the BMS h the event that a high VCTlevel exists. A failure in the line segment wiH not result in an initiating event. The failure can be detected based on low levelin the i
holdup tank and area radiation alarms, and then isolated. Similar to the other segments in the letdown line, this segment is not needed to accomplish or support any of the safety functions following plant shutdown.
A LOW consequence category would therefore be assigned to the failure of this segment. No degradation mechanisms were identified for this line segment. Based on the LOW consequence D
category and no leak potential, the risk significance due to the segment failure would (G
be LOW (i.e., CAT 7).
Since no element selections are needed for low risk-significant segments, the welds for this line were not entered in the database.
3.2.4 Shutdown purification return line from downstream of valve 2CVC 146 to upstream of valve 2SI 35 (2HCB-179 3*)
This line segment provides a return flow path for the reactor coolant that is diverted to the letdown portion of the CVCS during shutdown operations. During shutdown cooling purification, a small portion of reactor coolant is diverted from downstream of the LPSIpump discharge to the letdown purification filters and ion exchangars in order to maintain the water chemistry of the RCS. During normalpower operation, the segment is normally isolated and a failure would nct cause an initiating event.
The line segment is not required to accomplish or support any of the safety functions needed to mitigate a design basis event. Thus, the resulting consequence would be LOW. Shutdown purification is manually initiated during the latter stnes of shutdown cooling operation when RCS temperature and pressure are at refueling conditions.
Because of the manual alignment that is required for shutdown purific= tion, a failure would be detected and isolated. Since no significant impact on shutdown cooling operation is expected, a LOW consequence is also assigned. No degradation mechanisms were identified for the above line segment. Based on the LOW consequence and no leak 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 this line were not enteredin the database,.
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ABB Combustion Engineering Nuclear Operations
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Calculation No. A PENG-CALC 014, Rev. 00 Page 9 of 53 3.2.5 Boration flow path from downstream of check valve 2CVC 63 to the VCT (2HCC.
76-3', 2HCC 76-2', 2HCB-57-3')
These line segments provide makeup to either the VCT or the suction of the charging pumps. The makeup is provided from the discharge of the boric acid makeup pumps, the reactor makeup water discharge pumps, or the desired combination from both pumps. The flow path is used for boron addition / dilution during normalpower operation on an as needed basis. This flow path is aligned to allow makeup to go directly to the charging pump suction when needed. Makeup to the VCT requires local manual action to reposition the valve in the flow path. A failure in any of the above line segments would result in the loss of this makeup path. Other makeup paths to the charging pump suction would not be affected.
Depending on the break location, draining of the VCT may also occur. This would cause a low boric acid makeup pump discharge pressure to be annunciated in the control room. Because of the control room annunciation, it is highly probable that the operators would secure the boric acid makeup pumpfs) and close the reactor water makeup and boric acid makeup valves. A segment failure is not expected to cause an initiating event. Due to the high probability of detecting and isolating the failure, and the availability of multiple attemate paths for boration, the resulting consequence is LOW.
No degradation mechanisms were identified for these segments. Based on the LOW consequence category and no leak potential, the risk-significance due to a segment failure would be LOW (i.e., CA T 7). Since no element selections are needed for low risk-significant segments, the welds for these lines were not enteredin the database.
3.2.6 Charging line from downstream of VCT to upstream of RCS charging line isolation valves 2CV-4827-2 and 2CV-48312 (2HCB-46-4", 2HCB-2-4", 2HCB-2 2", 2HCB-67-2", 2HCB 68 2", 2HCB 17 3", 2HCB-33-3", 2CCB-8-3", 2CCB 112", 2CCB-2S-2', 2CCB-9 2")
These line segments provide makeup for the diverted reactor coolant that occurs during the letdown process. To maintain the pressurizer level (i.e., RCS inventory) within the desired operating range, the makeup compensates for the totalletdown flow and the controlled RCP bleed-off flow. A failure in any of the above line segments would result in a loss of charging.
There are several control room irdications or alarms to identify the loss of charging, including low charging header flow, low charging header pressure, or a mismatch between the charging header pressure and RCS pressure. Because of the direct control room indications and alarms, it is highly probable that the failed segment would be detected and isolated followed by the initiation of a controlled plant shutdown. An alternate charging flow path via HPSIheader #1 is available if the break occurs downstream of the chbrging line isolation valve 2CV-4840-2. In addition, the HPSI system can be used to provide RCS makeup once the RCS pressure is reduced below the shut-off head of the HPSI pumps following plant shutdown. Because a controlled plant shutdown would be initiated (i.e., design basis category ll event), a LOW consequence category is assigned due to a segment failure.
The potential of a small LOCA exi,.ts if the segment failure occurs inside the containment and both the check valve and motor operated valve fail to close. The combined effect of the conditional core damage probability for small LOCA and the ABB Combustion Engineering Nuclear Operations
A R It D%lFEF Calculation No. A-PENG CALC-014, Rev. 00 Page 10 of 53 probability of both valves failing to close results in a LOW consequence.
The potential for containment bypass also exists if a failure occurs outside the containment and downstream of the charging header isolation valve. The active barriers (i.e., check valve and motor-operated valve) must fail in order for the containment to be bypassed. The impact on containment performance results in e MEDIUM consequence. A MEDIUM consequence is therefore bounding for these segments. No degradation mechanisms were identified for these charging line segments. Based on the MEDIUM consequence category and no leak potential, the risk-segment failure would be LOW (i.e., CAT 6). Since no element selections are needed for low risk significant segments, the welds for these lines were not entered in the database.
3.2.7 Boration lines from downstream of either boric acid makeup tank to the associated boric acid makeup pump (2HCB 9-4", 2HCB 9 3")
These lines align the boric acid makeup pumps and the gravity feed boration line to the concentrated borated water in the associated boric acid makeup tank. A failure in any of the above line segments would result in depletion of boric acidinventory in the associated boric acid makeup tank and a resulting low level condition in the affected tank. The low level condition is indicated and annunciated in the control room. Because of the direct annunciation in the control room, there is a high probability that the failure will be detected and isolated if possible within a short time after the occurrence.
There are several flow paths available for providing
(
borated water to the RCS, including (1) flow path from the unaffected boric acid t
makeup tank via the associated makeup pump and charging pumps, (2) flow path from the unaffected boric acid makeup tank through the gravity feed line and charging pump, (3) flow path from the RWT and charging pump. Because the failed segment would not cause an initiating event and at Icast two backup flow paths are available for non-anticipated transients without scram (ATWS) events, a LOW consequence category would be assigned.
The ANO 2 IPE (Reference 9.16) assumed that baration is required to prevent core damage resulting from an A1WS scenario caused by mechanical failure of the rods to insert into the core, and the probability of failing to inject borated water is l
approximately 1.0E-02. The above segments are considered as part of the injection flow path. Of the three initiating events (turbine trip, loss of main feedwater and loss of o; alte power) considered followed by failure to scram, ATWS scenarios involving turbine trip were the most risk significant. According to the IPE, the conditional cre damage probability of the failed segment (i.e., failure to provide 7
boron injection; is approximately 2.0E-5.
Thus the resulting consequence is MEDIUM, which is bounding for these segments.
No degradation mechanisms were identified for the line segments. Based on a MEDIUM consequence category and no leak potential, the risk-significance due to the failed segment would be LOW (i.e., CAT 6). Since no element selections are needed for low risk-significant segments, the welds for these lines were not entered I
in the database.
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ABB Cornbustion Engineering Nuclear Operations
bMIFIN Calculation No. A PENG CALC-014, Rev. 00 Page 11 of 53 3.2.8 Boration lines from downstream of either boric acid makeup pumps to upstream of check valve 2CVC 58 (2HCB 14-3', 2HCB 25 2", 2HCB-B12")
These lines provide emergency boration via the boric acid makeup pumps. A limiting break associated with the above line segments would prevent emergevy boration via the boric acid makeup pump path. A low boric acid pump disc.iarge pressure would be indicated and annunciated in the control room during the operation of the pump. Because of the direct control room annunciation, there is a high probability that the boric acid pumps would be secured to terminate flow through the break.
The redundant boration path via the gravity feed line would not be affected. In o
addition, the boration flow path from the RWT would not be effected. Because the failed segment would not cause on initiating event and there are at least two backup flow paths available, a LOW consequence category would be assigned.
No degradation mechanisms were identified for these lines.
The above segments are also considered as part u:.he boron injection flow path for mitigating an ATWS.
Similar to the segments in Section 3.2.7, the resulting consequence is assigned as MEDIUM which is bounding for these segments.
No degradation mechanisms were identified for these lines. Based on a MEDIUM consequence category and no leak potential, the risk significance due to the failed segment would be LOW (i.e., CAT 6). Since no element selections are needed for low risk-significant segments, the welds for these lines were not entered in the database.
3.2.9 Gravity feed boration line from downstream of gravity feed valves 2CV-49201 and 2CV-4921-1 to upstream of check valve 2CVC-49 (2HCB-60-3')
The segments of this line provide a boration flow path from the boric acid makeup tanks to the RCS via the charging pumps. The segment is normally isolated during power operation, and a failure will not cause an initiating event. However, the gravity feed boration path will become inoperable. A low charging pump discharge pressure may occur to indicate ths segment failure. Once the failure is detected, the operator can close the gravity feed valves from the control room. Boration using the boric acid makeup pump paths will not be affected. Following a LOCA, the ECCS pumps are used to inject borated water into the RCS once the RCS pressure drops below the shut-off head of the ECCS pumps. Since the boric acid makeup pumps and the ECCS pumps are available for delivering borated water to the RCS following a LOCA (i.e., a design basis category IV event), a LOW consequence category would be assigned to the failed segment.
The above segment is also considered as part of the baron injection flow path for mitigating ATWS.
Similar to the segments in Section 3.2.7, the resulting consequence is assigned as MEDIUM which is bounding for this segment.
No degradation mechanisms were identified for these line segments. Based on a MEDIUM consequence category and no leak potential, the risk-significance due to the failed segment would be LOW (i.e., CAT 6). Since no element selections are needed for low risk significant segments, the welds for this line were not entered in the database.
1 ABB Combustion Engineering Nuclear Operations
R It R MIFIF Calculation No. A-PENG-CALC 014, Rev. 00 Page 12 of 53 3.2.10 RWT fiH line from downstream of valve 2CVC 64 to the RWT (2HCB-613', 2HCB-30 3')
These line segments provide the floa path for fi/Eng the RWT with the required concentration of boratad water. The fining of the RWTis performed during refueling or shutdown operations. The segments are normaHy isolated during power operation and a failure wiH not cause an initiating event. These lines are also not needed to support cr accomplish any of the safety functions for mitigating a design basis event. Because the segments are manuaHy aligned for fiHing the RWT, a failure is considered to be detectable and isolable. The resulting consequence is assigned as NONE. Based on this consequence category, the risk significance 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.
3.2.11 Fuelpoolfinline from downstream of valve 2CVC 66 to the fuelpool(2HCC-75-3')
This line segment provides the flow path for fiHing the fuel pool with the required concentration of borated water. The segment is normaHy isolated during power operation and a failure would not cause an initiating event. The fiHing of the fuel pool is performed when the RCS temperature and pressure are at refueling conditions.
Because the segment is manuaHy aligned for adding concentrated borated water to the refueling pool, a failure is considered to be detectable and isolable.
The failure win have no impact on shutdown cooling operation.
The resulting consequence is therefore assigned as NONE. Based on this consequsnce category, the risk significance is LOW (i.e., CAT 71. Since no element selections are needed for low risk significant segments, the welds for this line were not entered in the database.
3.2.12 Lines with NominalDiameter 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 conduction of a system leakage test. Consequently, since this piping is already subject to system leakage testing by the ASME Code, a risk assessment of this piping is not warranted.
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Calculation No. A.PENG. CALC 014, Rev. 00 Pope 13 of 63 TABLE 1 CSTS Hol/NDARIES Dnr bne Lksmptom 15 0 Pope CoJr Pure Nomsnal Number Dntwung Class Dsameter nn) 2CCA.]2 2*
Lesdown bne. Upstream ofRegenemtsve ils inIrt 2CCA.I21 1
2 l'alve 2CCA 26 2" Chargsng bne to RCS Coldleg 2P.32C XCA 2614 1
2 2CCA.2613 2CCA.26-16
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XCA 27 2*'
Chargung Une to RCS Cold Leg 2P.32B XC4 271 1
2 X CA.27 2 2CC4 27 3 XCA.17 4 l
XCB-l.2 "
Letdosm kne. 2"Fortsm Upsteram of 2CCB.I.!
2 2
Regenemttvr its 3CCB.I.2%"
Lt.doun bne. 2Ik"Ponson Upstream qf 2CCB.l.1 2
2.3 Rcrentrative lix XCB.).4 "
Letdown bne. 4" Portwn llpsturam of 2CCB l.1 2
4 Regenemtsve ils 2CCB.I8" Letdown bne. 5 "Ponwn Upstream of 2CCB.I.I 2
o Regenomitvr H t
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LetJoun bne.1" Ponom Downstream of 2CCB 21 2
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Calculation No. A PENG CALC 014, Rev. 00 Y
Page 17 of 53
- 4. 0 CONSEQUENCE EVALUA TION The Chemical and Volume Control System (CVCS) is interconnected with the Reactor Coolant System (RCS) to provide makeup for inventory lost during normal power operation, and to maintain reactor coolant chemistry and reactivity control. During normal power operation, a smaH omount of reactor coolant (normaHy 40 ppm) is diverted via the letdown line to letdown filters andion exchangers. The reactor coolant temperature and pressure are reduced by the regenerative and letdown heat exchangers, and by letdown and back pressure control valves prior to entering the purification filters and ion exchangers where insoluble particulate and ionic impurities are removed. The purified reactor coolant then flows to the VCT which serves as (1) a surge volume for the RCS, (2) a reservoir to ensure adequate NPSH for the charging pumps, (3) a means of maintaining the Ha concentration in the RCS, and (4) a meth0d for determining if RCS leakage exists. Taking suction from the VCT, the charging pumpfs) reium the letdown flow to the RCS in order to maintain the RCS inventory within the desiredlevels. During an emergency condition (e.g., a LOCA), both the letdown flow and VCT outlet flow are isolated.
The charging pumps win then be automaticaHy re aligned to the boric acid makeup tanks fot injecting con ~entrated borated water into the RCS. The consequence evaluation of the RCS is treated separately and is therefore not includedin the consequence evaluation of the CVCS.
The consequence evaluation for the CVCS was performed based on the guidance provided in the EPRIprocedure (Ref. 9.11. The evaluation focused on the impact of a pipe segment failure on the capability of the CVCS to perform its design functions, and on the overell Q
operation of the plant.
Impacts due to direct and indirect effects were considered.
GeneraHy, the effects of a direct impact are LOCA initiating events. An indirect effect resulting from the faHure of a pipe segment would be the impact on neighboring equipment within the CVCS or other interfacing systemis). Indirect impacts would generaHy 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 major equipment of the CVCS considered in this evaluation is located inside the containment. Certain safety related components which are designed to mitigate a LOCA are also located inside the containment. In general, the spatial effects of a segment failure are primarily associated with flooding, spraying, or jet impingement. Dynamic analyses (Ref.
9.3, Sections 3.6.4.2.9 and 3.6.4.2.10) have been performed to assess the spatial effects of failed CVCS piping. These analyses have concluded that a failure in the CVCS piping wiH not cause a more severe design basis event than the LOCA initiating event itself. Depending on the break location, the redundancy of certain engineered safety features needed to mitigate the pipe failure may be sfiected. However, the ability of the engineered safety features to mitigate the consequences of a LOCA wid not be impaired.
The environmental effects caused by a failure of the CVCS piping have also been assessed.
AH safety injection components located inside the containment have been designed to withstand the LOCA environment (SAR Section 6. 3. 2. 1 2. 11. The containment cooling units are also designed to maintain their functionalintegrity in a post LOCA environment (SAR Section 6.2.2.2.2).
A system walkdown was conducted as part of the consequence evaluation. The walkdown revealed that the line segments outside the containment which are entered in the database
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ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 014, Rev. 00 Page 18 of 53 are located in the Upper South Piping ana Penetration Area of the Reactor AuxHiary Building (RAB). These line segme~ts include the letdown line from downstream of the containment penetration to upstream J the letdown control valves. The foHowing was observed while conducting the walkdown.
(a)
The Upper South Piping and Penetrat!on area is located at elevation 360' O'.
There are two drains within this area.
(b)
Entrance to this area is through a non water tight door. It is judged that the force exerted on the door vriH cause it to open and provide a propagation path to the adjacent flood tone (i.e., RAB 2078 DD). The en'rance door wiH be forced open before a significant amount of water can accumulate inside the room and flood the valve motors.
The outflow of water win propagate to lower elevations and eventuaHy to the RAB sump via open grating, stairway No. 2001, and the floor drain 1
system.
(c)
There are no motor control centers or switchgear within this area.
(d)
The HPSI, LPSI, and CS line infection volves are located in this area. Other safety related valves located in the area include the HPSI hot leg injection valves, the servke water firae isolation valves for containment cooling units 2VCC 2A and 2VCC 28, the shutdown cooling line isolation valve, and the ERV distribution valves to steam generator 2E 24A.
(e)
The limitorque actuators for the motor operated valves are sealed to minimize the effects caused by spraying orjet impingement.
The majority of the line segments not includedin the database are located in Tank and Pump rooms and Corridors flood tone (i.e., RAB 2040 JJ) of the Reactor Auxiliary Building. In addition to the CVCS components, the service water header valves, RWT discharge valve and motor control center 2B52 are located in this flood zone. This motor control center provides power to safety related valves. During the walkdown, it was also observed that the above components are located sufficiently far from the postulated break location such that they would not be impacted. It was also observed that this flood zone occupies a large area and includes several floor drains. Because of the large area, the accumulation of a significant amount of standing water to threaten the operability of the above motor control center is unlikely before corrective actions are taken. Hence, the impact of spatial effects is considered to be insignificant.
A walkdown of the CVCSline segments inside the containment was not performed. At the time this evaluation was being performed ANO-2 was operating et normalpower. Access to the containment was therefore not feasible. The dynamic analyses of a failure in the CVCS piping and the assessment of engineered safety features equipment capability to withstand the post LOCA environment have concluded that the mitigation of a LOCA wiH not be impaired.
he dynamic analyses and the post LOCA environment assessment ha.re 7
addressed the spatial effects of a CVCS pipe failure inside the containment. It is therefore believed that a walkdown of the CVCS would provide no new insights regarding the impact of spatial effects and is not needed.
S l
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ABB Combustion Engineering Nuclear Operations l
ABB Calculation No. A PENG CALC 014, Rev. 00 Page 19 of 53 In performing thir evaluation, several types of Input were used and several assumptions were made.
These inputs and assumptions are discussed in Section 4.1.
Eight consequence segments were identified for the CVCS lines entered in the database. Of the eight, two were assigned as HIGH, five as ' MEDIUM
- and one as ' LOW'.
The consequence assessment summary for these segments is provided in Section 4.2.
The bases and justifications for each category assignment are provided in Appendix A.
This appendix contains reports obtained from the ISIS software (Reference 9.2) for the CVCS.
For the CVCS lines not entered in the database, four were assigned as ' MEDIUM
- consequence, five as
- LOW' consequence and two as 'NONE'.
No leak potential was identified for these line segments.
4.1 CONSEQUENCE ASSUMPTIONS 4.1.1 It is assumed that the pipe degradation process is relatively slow and that pipe failure tends to occur randomly in time.
The exception to this assumption of randomness is for certain piping segments which are seldom exposed to the normal operating pressures of the CVCS considered in this evaluation. In these cases, it is assumed that the piping is so weakened by the degradation process that upon demand there would be a sudden failu.~. For portions of the CVCS where the piping is normany or regularly exposed to opeo. sting RCS pressures, it is assumed that the piping failure is most likely to cause a de, tion basis LOCA. SpecificaHy, it is assumed that:
(a)
Piping downstream of the charging line check valves can only failin such a G
way as to cause a smaH LOCA initiating event and also to cause the unavailabHity of charging via this path after the LOCA.
(b)
Piping in the letdown line which cannot be isolated from the RCS can only failin such a way as to cause a smau LOCA initiating event and also the loss of letdown.
(c)
Piping that is downstream of the normaHy closed drain valves in the letdown and charging lines inside the containment is usuaHy not exposed to RCS operating pressures. For this piping, it is assumed that the faHure occurs during a demand or if the potential for exposing the segment to RCS operating conditions exists.
- 4. t.2 CVCS pipe segment failures that cause a LOCA are assumed to occur during full power operating condition of the plant.
4.1.3 Because the nominaldiameter of the CVCS piping of concem is two inches or less, a segment failure in the piping which cannot be isolated from the RCS is assumed to cause a smaH LOCA. This is based on the ANO 2 IPE (Reference 9.16, Tables 3.1 1
& 3.1 3).
4 1.4 Based on the dynamic analyses of CVCS pipe breaks and the capabHity of mitigating components to withstand a LOCA environment, it is assumed that spatial effects of a segment failure in the RCS piping is negligible or of no concern. Closure of the (3
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)
assumed foHowing the generation of an SIAS.
ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 014, Rev. 00 Pope 20 of 53 4.1. 5 The letdown control valves and charging pumps are configured to operate in the automatic mode.
4.1.6 A potential LOCA Initleting event was considered for pipe segments normally Isolated from the RCS (i.e., charging and letdown drain lines). A potential LOCA is defined as the failure of the isolating valve followed by the failure of the associated CVCS segment. Since the dialn lines are of 2' nominal diameter, the resulting consequence for a potential LOCA is based on the faHure potential of the isolating valve and the Conditional Core Damage Probability (CCDP) for a small LOCA.
Specifically, la)
The failure probabHity of the upstream manuel valve in the drain line is approximately 4.0E 2 (Reference 9.16). The combined effect of the foHure of the manual valve and the CCDP for sman LOCA (Table 1 of Appendix A)
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results in a MEDIUM consequence.
j (b)
By including the failure potential of the second Isolation valve in series for any of the drain lines, the resulting consequence would decrease from a MEDIUM to a LOW. Two valve failures must occur in order to expose the piping to a potentialLOCA.
(c)
A segment failure in the charging line and failure of the associated downstream check valve to close wiH also result in small LOCA.
The probability of the check valve failing to close is approximately 2.0E4 (Reference 9.16). The combined effect of the check valve failure and the CCDP for small LOCA (Table 1 of Appendix A) results in a LOW consequence.
O ABB Combustion Engineering Nuclear Operations 4
~
Calculation No. A.PENG CALC 014, Rev. 00 Page 21 of 53 4.2 CONSEQUENCE IDENTIFICA TION The consequence summary assessment is provided in tabular form in this section. Table 2 summarises the consequence evaluation for the CVCS. This table contains the following information for each of the consequences identified:
ConsequenceID A unique number assigned to the consequence Boundary The figure number that illustrates the boundaries for the consequence Description A brief description of the effects of the consequence DB Event Category The category of design basis initiating event caused by a failed pipe segment of the CVCS, 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 mitigating system (s) or train (s)
Available Trains The number of trains available for performing the intended design function of the mitigating systems Consequence Cat.
The assigned consequence category based on the application of the methodology providedin the EPRIprocedure (Ref. 8.1)
G Simplified schematics are provided in ficu'es 2 and 3 to illustrate the boundaries for each of the CVCS consequences.
Dotted lines are used to identify the boundaries for each consequence. Major CVCS equipment is shown on these figures for esse ofidentification.
The bases and justifications for each of the assigned consequences are documented in Appendix A.
The ISIS tref. 8.2) software was used as a tool to prepare the documentation in this appendix. The documentation of the spatial effects is besed en a review of the dynamic analyses provided in the ANO 2 Safety Analysis Report (Ref. 8.3i.
Table 3 provides a cross reference to the CVCS consequences, their corresponding figure numbers and the corresponding Isometric Drawings.
4.3 SHUTDOWN OPERA TION AND EXTERNAL EVENTS Shutdown Operation The consequence evaluation is an assessment assuming the plant is at-power. Generally, the at-power plant configuration is considered to present the greatest risk for piping failures since the plant requires immediate response to satisfy reactivity control, heat removal, and inventory control. By satisfying these safety functions, the plant will be shut down and maintained in a stab!e 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 (Reference 9.1) provides no guidance on consequence evaluation during shutdown operation. This limitation is assessed herein to gain some level of confidence that the consequence ranking during shutdown would not be more limiting, n
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v ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC 014, Rev. 00 Page 22 of 53 Pipe segments that are already ranked as 'HIGH' consequences from the evaluation at-power need not be evaluated for shutdown. Those that are already 'A4EDIUA4' require some confidence that 'HIGH' would not occur due to shutdown configurations. However, the ' LOW' consequs..ces for power operation requires more confidence that a 'HIGH' would not occur and some confidence that a *A4EDIUA1' 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 results of the comparison indicate that during shutdown cooling operation certain line segments of CVCS are used to provide RCS purification.
However, during shutdown operation une RCS operating conditions are less severe than during power operation, and the appropriate line segments are aligned locoHy in order to purity the RCS when required. A segment failure is therefore considered to be detectable and isolable. Because of these conditions there wiH be no significant impact on shutdown cooling operation, thus the resulting consequence is LOW. The consequence ranking during shutdown operation would not be more significant than the ranking at power.
External Events Although external events are not addressed in the current version of the methodology (Reference 9.1), the potential importance of piping faHures during external event is also considered.
The ANO 2 IPEEE was reviewed to determine whether external initiating even i, with their potential common cause impacts on mitigating systems, could affect consequence ranking. This information, along with information from other extemal event PRAs, is considered to derive insights and confidence that consequence ranking is not more significant during an extemal event. The fonowing summarizes the review for each of the l
major hazards (seismic, fire and others).
Seismic ChaHenges The potential effects of sei.;mic initiating events on consequence ranking is assessed by considering the frequency of chauenging plant mitigating systems and the potentiall. oact on the existing consertuence ranking. The foHowing summarizes this assessment:
GeneraHy, the CVCS piping considered in this evaluation has a seismic fragility capacity much greater than the 0.3g screening value and is not considered likely to i
fait during a seismic event.
Failure of CVCS piping during power operation already causes an initiating event.
The frequency of an earthquake induced pipe failure in the C\\'CS is less than the at.
power value assumed in the evaluation. Also, the likelihood of a simultaneous seismic event doshg or after a pipe break is low.
Reactivity control!s unlikely to be affected by seismic events because loss of offsite power (the most liksly scenario) wiH de-energize and drop controlrods. A very large earthquake may conse mechanical failure of the core and/or prevent rods from entering the core. However, such a low probability event would likely impact most functions due to equipment failures, causing core damage. The importance of pipe failure becomes irrelevant at this point and it is an extremely low probability event.
Based on the above, the consequence ranking for the CVCS during a seismic event is enveloped by the at power consequence ranking.
ABB Combustion Engineering Nuclest Operations
ABB Calculation No. A.PENG CALC 014, Rev. 00 Page 23 of 53 Fire Challenges ~ The ANO 2 (PEEE Indicates that the fire core damage frequency is dominated by fires initiated outside the containment. Fires are not assumed to cause a LOCA.
The consequence ranking for fire events are therefore enveloped by the consequence nt power.
Other External Challenges - Other hazards were screened in the ANO 2 IPEEE and are assumed to have little or no risk significant impact on the CVCS.
O O
i ABB Combustion Engineering Nuclear Operations
mob 2% WW Calculation No. A-PENG-CALC-014, Rev. 00 Page 24 of 53 Table 2 CVCS Consequence Assessment Summary C-avence D Bowdery Deservren DB Event Oweet Effects Speost Etfacts krpect Grossr AwateNo Witmetag Consequence Category Trairns Categewy CVCS-C-CTA figwe 2 loss of reector iw:.; wre W
SmeR L OCA irutmteg IUone Hsieteg event AR ECCS irrec3cer kops H!GH 2P32C chorgmg &ne occurs event eruf kss of during rwwmatpower operaten.
CVCS-C018 figure 2 Loss of reector cr0%-; via N
SmeE LOCA irwtretmg None kwesting everet AN ECCS & loops HIGH 2P328 chargesg Ene occurs event armikss of damp normerpower operation if_;,
CVCS-C-02 figwe 3 Loss of rea;- coolant wie IV SmeR LOCA irwtretmg None kweetmg event AR arevas of the HIGH fordown 6ne occurs during event erwtigetmg systems enema! power opereren.
CVCS-C43 figure 3 Nr ais.:kss of coolant une IV P&terrtief smeELOCA None None AR anins of the MOftlM RCS coki org 2P 32A &am Sne.
es;:2,tmeg systems CVCS C-04A figure 2 Pbtenoaf kss of c.-,; -.; wie IV hientief smeE LOCA None P&tenear AD weeas of the LOW 2P32C chargeg W.
irutiering event nwrigermg systems armf kss of cf-;g system CVCS-C-048 figure 2 P&tentief kss of coolent wie IV P&tentialsmet LOCA None 1%tentief AR trairns of the LOW 2P3 B chargmg Ene.
evemtmg event trwtigering systems eruf kss of
.f.yq system CVCS-C-OS
.We 2 Ptrtentia!kss of coolarrt wie IV P&tentief smet LOCA None 1%terrtial AR trohs of the MEDet/ht chorging Kne tro RCS caktleg hwtietmg everrt e.J;;, ;, y systems 2P3281 &ein 94ing CVCS-C-06 figwe 2 1%tentief kss ofiv..: ; vie IV PotenDalsmeE LOCA None ftstentief AR weens of the MEDIUM chargmg Ene (to RCS ccAfleg irwoetmg event erwtigating systems 2ft32Cl &ein p; ping-CVCS C-07 figure 3 Loss of k;,:vn. few ui.a5 R
isoleNe smet LOCA None kutietmg event AR Vans ofiN;.y. ^ ;
MEDIUM feaside contanmenti durvig entf has of systems norrnotpower operation.
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CVCS-C-08 figure 3 Loss ofletdown thw va--
R isolaNe smet LOCA None kwtieting everrt AM trens of..d;me;-,
MEDIUM loutskie.v..;-
-.;l during armikss cf systems normetpower operation.
ter -. -
ABB Combustion Engineering Nuclear Operations e
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Calculation No A.PENG. CALC.014. Rev. 00 Page 25 of 53 Table 3 CVCS Consequences, figures and Isoinetric Drawings ConsequenceID hpure Numtwr Isometric Drawings CVCS-CDIA 2
2CCA.2616 2CCA 2615 2CCA 2614 l
CVCS-C-018 2
2CCA 27 3 2CCA.27 2 2CCA 271 CVCS-C-02 3
2CCA.12 1 CVCS C-03 3
2CCA
- 12 1 CVCS C-04B 2
2CCA 27 4 2CCA 27 3 CVCS C-04A 2
2CCA 2616 CVCS C OS 2
2CCA.27 3 CVCS C-06 2
2CCA 2616 CVCS.C-07 3
2CCA 12 2 2CCB11 2CCB.2 1 CVCS-C-08 3
2CCB 2 2 r3O c.U ABB Combustion Engineering Nuclear Operations
AnR D'1B E Calculation No. A-PENG-CALC-014. Rev. OC' Page 26 of 53 i
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SCHEMA TIC OF CVCS LETDOV.W FLOWPATH TO LETDOWN CONTROL VALVES ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG. CALC-014, Rev. 00 Page 28 of 53
- 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 CVCS, 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:
ThermalFatigue (TF)
Thermal Stratification, Cycling, and Striping (TASCS)
Thermal Transients (TT)
Stress Corrosion Cracking (SCC) e Intergranular Stress Corrosion Cracking (IGSCC)
Transgranular Stress Corrosion Cracking (TGSCC)
External Chloride Stress Corrosion Cracking (ECSCC)
Primary Water Stress Corrosion Cracking (PNSCC)
Localized Corrosion (LC)
Microbiologically influenced Corrosion (MIC)
- Pitting (PIT)
- Crevice Corrosion (CC)
- Flow Sensitive (FS)
Erosion-Cavitation (E C) f'
- Flow Accelerated Corrosion (FAC)
(
In per4orming this evaluation, some basic inputs were used. These inputs are discussed in Section 5.3. The criteria andjustifications are providedin 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 only one degradation mechanism is potentially present: thermal fatigue. Using ISIS (Reference 9.2), two damage groups (DM groups) were identified as CVCS T and CVCS N, and are defined in Table 4 below. These DM groups result in two failure potential categories: "Small Leak" and *None'.
The FMECA. Degradation Mechanisms for each segment and associated elements are presentedin Appendix B.
Table 4 Damage Groups Demsee Demeee Mechenleme FeDwe Group Thermalfetieve Stress conoelon Crackine LoceRred Conoeron Row Seneeske Potendel Jo TAsCS TT IQsCC TGsCC ECSCC PWSCC MC PfT CC EC FAC Cetecorv i
CVCS T No Yes N2 No No No No No No No No SmeMleek CVCS N No No No No No No No No No No No None (vD 1
l ABB Combustion Engineering Nuclear Operations
ABB Calculation No. A PENG CALC Old, Rev. 00 Page 29 of 53 6.1 DAMAGE GROUPS 5.1.1 DM GROUP: CVCS-T The CVCS T damage group is considered subject to thermal fatigue due to the potential for thermal transients (TT). The effected sections are describedin the table below and depictedin Figures 2 and 3.
Une Number Affected Sections Thermalfatigue Cause 2CCA 26 2*
Charging Line 2P32C hortiontal section thermal transients from the second upstream elbow to the charging nottle at the coldleg 2CCA 27 2' Charging line 2P328 horizontal section thermal transients from the second upstream elbow to the chstging nottle at the coldleg 5.1.2 DM GROUP: CVCS N The CVCS-N damage group is not considered susceptible to any damage mechanism. This group includes (See Figures 2 and 3):
the entire letdown line (2CCA 12 2', 2CCB 12", 2CCB 12%', 2CCB 14',
2CCB 18', 2CCB 2 2', 2CCB 2 2A*) from the cold leg norrie to control valves 2CV 4816 and 2CV-4817 and the drain path downstream of manual valve 2RC.
4A; charging line 2CCA 26 2' from isolation volve 2CV-48312 downstream to the affected section described above and the drain path downstream of manual valve 2CVC 1188 and; charging line 2CCA 27 2* from isolation valve 2CV-4827 2 downstream to the affected section described above and the drain path downstream of inanual valve 2CVC 1186.
6.2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICATION The degradation mechanisms and criteria assessed are presented in Table 5.
O ABB Combustion Engineering Nuclear Operations
ABB q
Calculation No, A PENG CAI.C 014, Rev. 00 IO' Page 30 of 53 Table 3 Degradation Mechanism Criteria andSusceptible Regions I
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Criteria Susceptible Regions TF TASCS
-nps > ) inch, and nostles, branch pipe
-pipe segment has a slope < 43*from horizontal (includes elbow or connections, sqfe ends, see into a verticalpipe), and welds, heat afected
-potential existsfor lowfow in a pipe section connected 4 a lones (HAZ), base component allowing mixing ofhot and coldpuids, 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 coldpuids, or fotential existsfor convection heating in dead-endedpipe sections connected to a source ofhotpuid, or potential existsfor two phase (steam / water) pow, or potential existsfor turbulent penetratton in branch pipe connected to headerpiping containing hotpuid with high turbulentfow, and
-calculated or measured AT > $0*F and
-Richardson number > 4.0 TT
-operating temperature > 270*Ffor stal.dess steel, or operatmg temperature > 220*Ffor carbon steel, and
-potentialfor relatively topid temperature changes includmg p) i coldpuid imectoon into hot pipe segment, or hotpuid enjection into coldpipe segment, and AT
> 200*Ffor stainless steel, or AT
> 130*Ffor carbon steel, or
! JT
> 0 allowable (appitcable to both stainless andcarbon)
-evaluated o accordance with extstmg plant IGSCCprogram per austenitic stamless steel (11HR)
NRC Generit, i.etter 88-01 welds and HAZ IGSCC
-operatmg Irmperature > 200*F, and (PH7()
-susceptable material (carbon content 2 0 03396), and
-tensile stress (includmg residual stress) is present, and
-oxsyen or oxidating species are present OR
-operating temperature < 200*F, the attributes above apply, and
-Intisatmg contammants (e.g., thiosu{ fate,fuoride, chloride) are also requtred to be present TGSCC
-operating temperature > 130*F, and austenitic stainless steel
-tensile stress (includmg residual stress) is present, and base metal, welds, and
-halides (e g., fuoride, chloride) are present, or flAZ caustic (NaCll)is present, and
-oxsgen or oxiditing species are present (only required to be present in conjunctson w halides, not required w' caustic) pO ABB Combustion Engineering Nuclear Operations
ABB Calculation No, A PENG. CALC 014, Rev, 00 Page 31 of 53 Table 3 (cont %
Degradation Mechanism Celterla and Susceptible Regions g,,'g,',"f","
'E "
Celteria Susceptible Regions SCC ECSCC
-operating temperature > 130*F, and austentric stainless steel
-tensile stress ispresent, and base metal, welds, and
-an outside piping surface is withinfve diameters ofaprobable llAZ leak path (e g., valve stems) and is covered wtIH non-metallic Insulation that is not in compliance wsth Reg. Guide 1.36, or an outside piping surface is exposed to wettingfrom chloride bearing environments (e.g., seawater, brackish water, brine)
PHSCC
-piping maternalis inconel (Alloy 600), and nonles, welds, andHAZ
-exposed to pramary water at T > 620*F, and without stress relief
-the material is mill annealed and cold u orked, or cold worked and welded without stress relief LC MIC
-operating temperature < 130*F, and patings, welds, IL42.
-low or Intermittent $ow, and base metal, dissimilar
-pH < 10, and metaljoints (e.g., welds,
-prearnce4ntrusion oforganic material (e g., raw water system), or fanges), and regions water source is not treated w4tocides (e.g, refueling water tank) containing crevices PIT
-potential existsfor lowpow, and
-osygen or oxidating species are present, and
-inttiating contamanants (e g. puoride, chloride) are present CC
-crevice condition extsts (e.g., thermal sleeves), and
-operating temperature > 130*F, and
-oxtgen or oxsd sing species are present 15 EC
-operating temperature < 230*F, and pitangs, welds, HAZ, and
-pow present > 100 hrspr, and base metal
-velocity > 30p%. and
-(Pc PJ / AP < $
FAC, -evaluated in accordance with existing plant FA Cprogram perplant FACprogram O
ABB Combustion Engineering Nuclear Operations
ABB f
Calculation No. A PENG CALC Old, Rev. 00 Page 32 of 53 5.2.1 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.
5.2.1.1 Thermal Stratification, Cycling, and Striping (TASCS)
CEOG Task 866 IReference 9.8) assessed the significance of stress levels due to thermal stratification in the CVCS charging lines (2CCA 26 2' and 2CCA 27 2') and letdown line (2CCA 12 2'). Due to continuous flow conditions during normalpower operations, this system is considered stable, and therefore, these lines are not subject to thermal stratification.
5.2.1.2 Thermal Transients (TT)
An assessment of loss of charging and loss of letdown transient for each specific pipe section is provided in the following tables. The horizontal sections of charging piping near the cold legs, as defined in Section 5.1, and assessed in the first table below, are considered subject to thermal transients. The remaining pipe sections, assessed in the second through sixth tables, are either not subject to thermal transients, or the ATs experienced have been determined to be acceptable per either Reference 9.18 or Reference 9.19.
O (l
V ABB Combustion Engineering Nuclear Operations
ARR 9%WW Calculatiort No. APENG-CALC-014. Rev. 00 Page 33 of 53 Mping thw /Desenprion Steady State Transent Event Bormdery Start Ca.. "o*v.3 Rior Event Sensence Range of ATs Esponenced Sormdery Stop to Eventinit*etion initietor
-+
-+
End initiet
-+
Finef 2CCA-26-2" / Charging *C*
Normat Operating Loss of Charging Letdown Isolated Chargeg Hestored Letdown Pmno M 143i n/a 4331 2307 Upstream of 2nd Dbow Pipe Temperature Ape Temperature Rpe Temperature Pipe Temperature Pipe Temperature 410'F 553*F 553'F 120'F 410'F r
@ Wd W
- Upon loss cf charging the temperature of this piping section win be warmed above its normal operating temperature of 410* to a
'8 2CCA-27-2" / Charging ~B"
- Dependent upon event duration, letdown win be isciated Umm M N h r
wa w
e s
(wam M
w e
to Charging Nozzle @ Cold Leg
- p n sesman n o containment ambient and charging fiow passing through regenerative heat exchanger wiB not have been heated)
- Upon restoration of letdown, the charging flow passing through the regenerative heat exchanger wiB be heated and het 410' water wiR be iriected into this cool piping Normal Operating Loss of Letdown Charging isolated Charging Restored Letdown Anwd 2301 433T 4331 2 SOT Pipe Temperature Pipe Temperature Pipe Temperature Pipe Temperature Pipe Temperature 410'F 120*F 553*F 120*F 410'F l
- Upon loss of letdown, the charging flow passing tfwough the regenerative heat exchanger wiB not be heated and cooler 120' water wm be iriected into tfis hot piping that has a normal cperating temperature of 410' l
- Dependent upon event duration charging wiB be isolated and the temperature of this piping section wiR be warmed to a maximum of 553' due to turbulent penetration flow from the Cold Leg (553')
- Upon restoration of charging cooler 120' water will be iriected into tNs het piping (water in charging piping wiu have cooled to containment ambient and charging flow passing through regenerative heat exchanger wiB not have been heated)
- Upon restoration of letdown, the charging flow passing through the regenerative heat exchanger win be heated and hot 410' water win be irjected into this cool piping I
l ABB Combustion Engi.
ing Nuclear Operations
O O
O aoo
- 't B W Calculation No. A-PENG-CALC-01A Rev. 00 Page 34 of 53 Mpingline /Descripson Steady State Transert Event Boundary Start Condtions nior Event Sequence Range of ATs Expermaced Boundary Stop to Everet initiation Initietor
-+
-+
End haiset
-+
-+
Rnef 2CCA-26-2* / Charging *C" Normal Operating Loss of Charging Letdown Isolated Charging Pulwed Letdown Im;w ed 2301 Downstream of 2CV-4831-2 Mpe Temperature Roe Temperature Rpe Temperature Rpe Temperature Rpe Temperature n/a 0
290f 410*F 120'F 120'F 120'F 410'F Upstream of 2nd Eibow
- Upon loss of charging, the temperature of this piping section will cool from its normal operating icmw.iwe of 410' to a 2CCA-27-2* I Charging *B" Downstream of 2CV 4827-2
~
i
- Upon restoration of charging.120' water win be iriected into this piping (charging flow passmg through regeneratrve heat exdangs Umm M N DW will not have been heated)
- Upon restoration of letdown, the charging flow passing through the regenerative heat exchanger wB be heated and hot 410' water will be iriected into this cool piping Normal Operating Loss of Letdown Charging isolated Charging Restored Letdown Tui-c4 Rpe Temperature Rpe Temperature Mpe Temperature Rpe Temperature Rpe Tww.in 2301 n/a 0
2SOI 410'F 120*F 120'F 120'F 410'F
- Upon loss of letdown, the charging flow passeg through the regeneratrve heat exchanger will not be heated and cooler 120* water will be iriected into this hot piping that has a normal operating Im..w. tore of 410'
- Dependent upon event duration, charging wiB be isolated i
- Upon restora$on of charging.120' water will be insected into this piping (chargeg flow passmg through reg =neratrve heat exchanger win not have been heated)
- Upon restoration of letdown, the charging flow passing through the regeneratrve heat exchanger will be heated and hot 410' water win be iriected into this cool piping i
ABB Combustion Engineering Nuclear Operations
91WW Calculation No. A-PENG-CALC-014. Rev. 00 Page 35 of 53 Piipirsgline / Description Steedy State Transient Event Bormdery Start C&2ka > Prior Events % ~.- -
Range of ATs Espenenced Boundary Stop to Event fusitiation Initietor
-+
-+
End initief
-+
-+
Finer 2CCA-12-2* / Letdown Normal Operating Loss of Charging Letdown isolated Chargmg Hestored Letdown Puted Letdown Wozzle @ Cold leg Pipe Temperature Pipe Temperature Pipe Temperature Pipe Temperattae Pipe Temperature n/a O
n/a O
553'F 553'F 553'F 553'F 553*F
- Loss of chargog
- Dependent upon event duration, letdown will be isolated and the temperature of this pi ing section will remac essentially at its t
normal operstmg temperature of 553' due to ttrbulent penetration flow from the Cold ieg (553')
- Charging is restored
- Upon restoration of letdown. 553' water from the Cold leg will flow through this piping Normat Operating Loss of Letdown Charging isolated Chargrnp Putw d Letdown Restored Pipe Temperature lipe Temperature Pipe Temperature Pipe Temperature Pipe Tm.w.iore O
n/a n/a O
553'F 553*F 553'F 553'F 553'F
- Upon loss of letdow 1, the temperature of this piping section will remam essentially at its normal vya.i~.v temperature of 553' due to turbulent penetrat on flow from the Cold Leg (553')
- Dependent upon even duration, charging wi!I be isolated
- Charging is restored
- Upon restoration of letdow m. 553* we_s from the Cola Leg will flow through this piping ABB Combustion Engi, ing Nuclear Operations
[d
\\
}
V v
ARR MWW Calculation No. A-PENG-CALC 41A Rev. 00 Page 36 of S3 Pl>mg line / Description Steady State Transsent Event Boundary Stort CondFrions Prior Event Sequence Menge of ATs Experienced Boundney Stop to Event Inrbetf>n trutsetor
+
+
End inisint
+
+
Final 2CCA-12-2* / Letdown Normal Operating Loss o' Charging Letdown isolated Charging Restored Letdown Restored 4331 n/a 433t
- '{
n/a Downstream of 3rd Dbow Rpe Temperature Rpe Temperature Rpe Temperature Rpe Temperature Rpe Tempe ature 553'F 553*F 120'F 120*F 553*F 2CCB-1-2* 4". and 8*
- tass of charging Intermediate piping between
- Dependent upon event duration, letdot-ft ill be isolateti and the temperature of this piping secten will cool from its normal operetng boundary start and stop poets temperature of 553' to a contamment <bient temperature of 120'
- Charging is restored
- Upon restoration of letdown hot 553' water from the Cold Leg will flow through this cool piping 2CCB-1-2 %
- Normal Operating Loss of Letdown Charging isolated Charging Restored Letdown Restored 4331 8
n/a n/a 4337 Inlet of Regenerative Heat Rpe Temperature Rpe Temperature Pipe Temperature Rpe Temperature Mpe Temperature Exchanger 553*F 120*F 120*F 120'F 553'F
- Upon loss of letdown, the temperat're of this piping sectron will cool from its normal m.i; s im.g.. tore of 553' to a contamment ambient temperature of 120*
i
- Dependent upon event duration, charging will be isolated
- Charging is restored IJpon restoration of letdown, hot 553' water from the Cold Leg will flow through this cool piping ABB Combustion Engineering Nuclear Operations
l
- 't W W Calculation No. A-PENG-CALC-014. Rev. 00 Page 37 of 53 Mping Une / Description Steady State Transient Event Boundery Start Cordtions PWor Event Sequence Roospe of ATs Experkxed Boundery Stop to Event initiation listietor
-e
-e End Ird6ief
-e
-e final 2CCB-2-2 % ~ / Letdown Normal Operating Loss of Derging Letdown Iso'ated Charging Restored Letdown Restored 3501 Outlet of Regenerative Hx Pipe Temperature Rpe Temperature Pipe Temperature Pipe Temperature Pipe Temperature 250T n/a 100T 220*F 470*F 120*F 120*F 220'F n,
m m
rer.
2CCB-2 - Upon loss of charging, the letdown flow passing through the regenerative heat exchanger wiB not be cooled and hot water at a Upstream of Penetration 2P14
- Dependent upon event duration, letdown will be isolated and the temperature of this prpeg section will cool from a maximum temperature of 470' to a cw6... eat ambient temperature of 120'
- Cnarging is restored
- Upo t restoration of letdown, warm water at a normal operating temperature of 220' wiB flow through this coci ppng NorrrJ Operating Loss of Letdown Charging isolated Charging Restored Letdown Li cf 1001 n/a n/a 100T Pipe Temperature P5pe Temperature Pipe Temperature Pipe Temperature Rpe Temperature 220*F,,,
120'F 120'F 120' 220*F m
,er.
- Upon loss of letdown, the temperature of this piping section will cool from its normal operating temperature of 220' to a contaenment ambient temperature of 120'
- Dependent upon event duration, charging wiu be isolated
- Charging is restored
- Upon restoration of letdown, warm water at a normal operating temperature of 220* wiB flow through this cool piping II) A normal operstmg termereture of 220* is assumed beoed on e review of typecal r+cu.:..; heet exchanger outlet termoretures recorded by the plant computer for temperature element 2TE-4820. This temperature occure durmo noemel letdown flow (v 4.;.e - (-40 gaml-
- 12) A mensmum temperature of 470* is concedered since, if the regenerstrve beet enchenger letdown outlet temperature se sensed by 2TE-4820 exceede 470*. then the letdown fm step veh.e 2CV-4820-2 will automaticafly close.
I i
i ABB Combustion Engi - ing Nuclear Operations
m n
(V,.)
(V)
(V) q l
- t 9% WW Calculation No. A-PENG-CALC-014. Rev. 00 Page 38 of S3 Mping Line / Description Steedy State Transient Event Boemdery Start C
".^;vai Prior Event Sequence Renge of ATs E>rpenenced Bosmdery Stop to Everre initiation Ermitiator
-e
-+
End heitief
-e
-+
Rnel 2CC8-2-2* / Letdown rJormal Operating Loss of Charging Letdown isolated Charging Restored Letdown Restored 3901 Downstream of Penetration 2P14 Rpe Temperature Pipe Temperature Rpe Temperature Rpe Temperature Rpe Temperature 250i n/a 140I 220*F "'
470*F "
E9'F" 80*F "
220*F "'
Upstream of 2CV-4816 / 4817
- Upon loss of charging, the letdown flow passing through the regenerstrve heat exchanger wiB not be cooled and hot water at a maximum temperature of 470' will flow through this piping that has a nonmat operating temperature of 220'
- Depe ident upon event duration, letdown wiB be isolated and the temperature of this pepmg secdon will cool from a maximum temperature of 470* to an auxiliary txelding smbient tw.m.iure of 80'
- Charging is restored
- Upon restoration of letdown, warm water at a normal w.b..v temperature of 220' will flow ttwough this cool ppng TJormal Operating Loss of Letdown Charging Isolated Charging Restored Letdown Restored 1401 Mpe Temperature Rpe Temperature Rpe Temperature Mpe Temperature Rpe Temperature n/a n/a 140i 220*Fn, 80'F 80'F m 80*F 220*F m
m n,
- Upon loss of letdown, the temperature of this piping secben will cool from its normal operating temperature of 220' to an auxiliary building ambient temperature of 80*
- Dependent upon event duration, charging will be isolated
- Charging is restored
- Upon restoration of letdown, warm water at a normal operating temperature of 220' wiB flow through this cool piping
- 11) A normel operating termereture of 220* is eseumed beoed on a review of typecal regeneestive heet euchenger outlet termeretures recorded by the plant cormuter for termeregueo element 2TE-4820. Thie temperature occure dunng normal letdown flow condersons I-40 gpml.
- 12) A mammum temperature of 470* is considered emee, if the regenerstrve heet orchenger letdewn evtset termere*ure se sansed by 2TE-4820 ene ede 970*, then the le down line stop a
velve 2CV-4820-2 wiB outomatically cicee.
C) An oundsery buildmg ternperature of 80* is eseumed beoed on a review of termeretuees receeded by the plant corrputer for venove temperature elemante located on stagnent system pipmg.
I ABB Combustion Engineering Nuclear Operations
I ABB Calculation No. A.PENG CALC Old, Rev. 00 Page 39 of 53 S.2.2 Stress Corrosion Cracking (SCC)
The electrochemicalreaction caused by a corrosive or oxygenated medium within a piping system can lead to cracking when con Ained with other factors such as a susceptible material, temperature, and stress. This mechanism has several forms with varying attributes including intergranular stress corrosion cracking, transgranular stress corrosion cracking, extemal chloride stress corrosion cracking, and primary water stress c~orrosion cracking.
S.2.2.1/ntergranular Stress Corrosion Cracking (IGSCC)
The piping warrsnting detailed consideration in this system operates in excess of the IGSCC temperature threshold of 200*F. However, none of this piping is exposed to oxygen or oxidizing species (the Volume Control Tank which serves as the primary charging supply source is maintained essentially oxygen free) and is therefore not considered susceptible to IGSCC.
S.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.
S.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 cc ssequently the potential for ECSCC to occur does not exist.
S.2.2.4 Primary Water Stress Corrosion Cracking (PWSCC)
PWSCC is not aoplicable as a potentialdamage mechanism for the CVCS due to the fact that there is no inconel(Alloy 600) present in the system.
S. 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.
S.2.3.1 Aficrobiologically influenced Corrosion (A1/C)
The piping warranting detailed consideration in this system operates in excess of the AflC upper temperature limit of 150*F. Consequently, this piping is not considered susceptible to A1/C attack.
O ABB Combuation Engineering Nuclear Operations
0 Calculation No. A PENG CALC 014 Rev. 00 O
Page 40 of 53 S.2.3.2 Pitting (pit)
The piping in this system is not exposed to oxygen or oxidizing species during normal power operation Ithe volume control tank which serves as the primary charging source is maintained essentially oxygen free) and plant chemistry controls ensure that initiating contaminants (e.g., fluoride, chloride) levels are negligible.
Consequently, this system is not considered sus Jtible to pitting attack.
l S.2.3.3 Crevice Corrosion (CC)
Thermal sleeves, which are considered prime crevice region locations, are installed in the charging nozzles (Reference 9.7). However, the piping in this system is not exposed to oxygen or oxidizing species during normalpower operation (the volu,,re control tank which serves as the primary charging source is maintained essentially oxygen free). Consequently, this system is not considered susceptible to crevice corrosion.
S.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 materialleeding to a reduction in wall thickness. Mechanisms that are flow sensitive, and can create this form of degradation include erosion cavitation and flow accelersted corrosion.
f3 O
S.2.4.1 Erosion Cavitation (E C)
Most of the piping warranting detailed consk'etation in this system, with the exception of the letdown piping downstream of the regenerative heat exchanger, operates in excess of the E C upper temperature limit of 250*F. This letdown piping however, which normally operates at a temperature of approximately 220*F, has a flow velocity less than 30 ft/s. Furthermore, there are no potential sources of cavitation (e.g., pressure reducing orifices or valves) in the piping warranting consideration. Consequently, this system is not considered susceptible to E C.
S.2.4.2 Flow f.ccelerated Corrosion (FAC)
The CVCS i: comprised entirely of austenitic stainless steel piping (Reference 9.4).
Since FAC is a phenomenon that only affects carbon steel piping, the CVCS is not susceptible to this degradation mechanism (Reference 9.12).
S.2. 5 Vibration Fatigue Vibration fatigue is not specifically made part of the EPRI risk informed ISI process.
Most documented vibrational fatigue failures in power plants piping indicate that they are restricted to socket welds in small bore piping. Most of the vibrational fatigue damage occurs in the initiation phase and crack propagation proceeds at a rapid rate t.nce a crack forms. As such, this mechanism does not lend itself to typicalperiodic inservice examinations (i.e., volumetric, surface, etc.) as a means of p
managing this degradation mechanism.
U ABB Cornbustion Engineering Nuclear Operations
M IDIF Calculation No. A.PENG-cal.C 014, Rev. 00 Page 41 of 53 Management of vibrational fatigue should be performed under an entirely separate program taking guidance from the EPRI Fatigue Management Handbook (Referenca 9.11). 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. Frequent system walkdowns, leakage monitoring systems, and current ASME Section XI system leak test requirements are some of the practical measures to address this issue. Because these measures are employed either singly or in combination for most plant systems it is not necessary to use a risk informedinspection selection process for vibration fatigue.
6.3 BASIC DA TA 5.3.1 All piping in the CVCS is austenitic stainless steel.
Under normal p.an:
operating conditions, the CVCS, as defined by the boundaries in Section 3.2, functions as indicated in the table below (References 9.4, 9.5, 9.6, and 9.17).
Mping Une NormalMont Operating Number Description Temperstwe Flow Condtlons 2CCA t2 2' letdown Une 553 Continuous Flow 2CCB t 2' letdown Line Upstream of 553 Continuous Flow 2CCB-t 2%
- Regenerative Heat Exchanger 2CCB t-4' 2CCB. t-8' 2CCB 2 2' letdr.svn Une Downstream of 220 Continuous flow 2CCB-2 2%
- Regenerative Heat Exchanger 2CCA 26 2' Charging Linss 4t0 Continuous Flow 2CCA 27 2' 5.S.2 Due to the cyclic nature of thermal transients, only those transients which occur d., ring initiating event Categories I and ll as described in Reference 9.1, Table 3.1 are considered in the evaluation of degradation mechanisms due to thermal fatigue. Category I consists of vase events which occur during routine operation, e tr., startup, shutdown, standby, refueling.
Cstegory ll consists of those events which have anticipated operational occurrer.ce, e.g., reactor trip, turbine trip, loss of feedwater. Therefore, the
)
transients.*o be evaluated are those transients which occur under norrnal operating ar.d upset conditions.
O ABB Combustion Engineering Nuclear Operations
Nb A'%IFIF l O Calculation No. A PENG-CALC 014, Rev. 00 Page 42 of S3
- 6. 0 SERVICE HISTORY AND SUSCEPTIBILITY REVIEW l
An exhaustive review was conducted from mid '96 to Spring '97 of databases (plant and industry) and station documents to characterire ANO-2's operating experience with respect to piping pressure boundary degradation. The results of this review are provided in a condensed form in Table 6 for the Chemical and Volume Control System.
Although several pre-commercial references are included for completeness, the timeframe for identifying items applicable to this effort was focused on post-commercial opstation (Commercial Operation date of March 26, 1980). This was done to avoid inclusion ofitems primarily associated with construction deficiencies as opposed to inservice degradation.
The foHowing 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 corre:ctive maintenance (CM) or modifications (MOD).
AdditionaUy, a separate query was performed in order to capture certain non-Code, O component failures.
This query tvas for non-Code Q and SR (safety related) components.
This database contains information from approximately 19M to the present.
- Condition Report (CR) Database The CR database was queried for any pipe leak / rupture events or other conditions associated with identified damage mechanisms at ANO-2.
The keywords searched under were; pipe, piping, la.e. water hammer, leak, leaking and leakage. CR's are written on Q, F or S equipment failures or other conditions potentially adverse to safety.
This database contains information from 1988 to the present.
- Licensing Research System (LRS)
The LRS databast' 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.
1"
$s ABB Combustion Engineering Nuclear Operations
b !k MIDIF Calculation No. A PENG CALC-014, Rev. 00
~Page 43 of 53
- Nuclear Plant Reliability Database System (NPRDS)
NPRDS was queried for ANO 2 entries fear pipe failures. The keywords searched under were: pipe. This database contains information from 1991 to the present.
ANO 2 ISIProgram Records The ISI program findings were compiled and reviewed for all outage and non-outage inservice inspections conducted at ANO 2 since commercial operation.
ControlRoom Station tog The station log was utilized as a source of information for recent operational events.
The log exists in electronic fo,m t from early 1994 to the present and has search capabilities which allowed a review for events of interest. The keywords searched under were: water hammer, leak and leakage.
- System Upper levelDocument (ULD)
The ULD was reviewed as a source for historical perspective of issue 3 related to the system and identification c! modifications made to the system or changes to operational procedures to address t a se issues (e.g., water hammer, corrosion or vibrational fatigue).
9
- Othar Station Documents This source of information consists of such documents as the SAR, Technical Specifications, operationalprocedures and the damage mechanism analysis done as part of this effort.
O ABB Combustion Engineering Nuclear Operations
,m f'
(b)
U
(
pS 1
(
Calculation No. A-PENG-CALC-014, Rev. 00 Page 44 of 53 Table 6 Service History and Susceptibility Review - Chetnicaland Volume ControlSystem Source Docsan.mte / Databasee Reviewed for Demare Mechereinme Ad&menecy Ceneidered Evidence ofIHetoricalfYpirsg Preenure Boundary Thermalfatigue Strees Corrosion Crocking locsEred Corresseur Row Senertive Mecherscal Water Other Degredation C. c. = st ANO 2 TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC VF Hammer Rndnge Station ksformation Manageme nt $ystem None None None None None None None None None None None None None None
.e=
Cormfirion Report Database None PEl1).
None None None None None None None None None None None I (,Si q "
Licensing Research System None None None None None None None None None None None
' (28 l '
None None Nuclear Plant Reliebnity Database System None None None None None None None None None None None Nsne None None 3 3 ANO-2 /S/ Program Records None None None None None None None None None None None None None
} J38 ' "
Contro/ Room Station Log None None None None None None None None None None None None None None System ())per leve/ Documents None None None None None None Nor e None None None None None None None Other Station Documents None
?P(4)l None None None None None None None
.None None None None None Leoend:
P (Precursor) - This category includes identification of postulated damage mechanisms and Icedings through knowledge of operstmg parameters, water chenestry, etc. No physicel evidence of pressure boundary degradetion currently existe. TNe category includes r.astulated. Tact...... identified es a result of tNo review.
PE (Plant Event) - This category includes identification of postulated damage nac:.s e...c and toedings as a reeutt of an obeerved or potential plant event (e.g., water hemmer). No physical evidence of pressure boundary degradation currently existe.
PD (Physical Damage) - This category meludse identification of observed pressure boundary degradation es evidenced by cracking, petting, westege, thinning, physical deformation or othat deterioration.
P8F (Pressure Boundary Failure) - This category includes identification of through-wall flaws resulting from the effects of en identified damage machenism.
Notes:
1.
Reference CR 2-92-0138 which documente the identifiestion and evefustion of charging system norrie thermal trenoiente.
2.
Reference LER's79-031, 80-019,80-034. 80-061,80-068, 80-090,81-001 and 82-007, all of which document small diameter (st " NPS) pipe or otherwise out-of-scope leskege primarily attributable to vibrational fatigue.
3.
Reference CR 2-94-0183, CR 2-954368 and ISI program records. Muttipie surface indications have been identified over twne in the CVCS. These indicatens were either removed or evefusted and deterrnined to be Code ecceptable. None of these indications were attributed to en inservice damese mechanism and are beisewed to have been norwervice induced (i.e.,
fabrication or other origin).
4.
Reference Section 5 of this document which oosntifies the potential for TT in specific portions of the CVCS.
ABB Combustion Engineering Nuclear Operations
M BRIN Calculation No. A PENG CALC-014. Rev. 00
'Page 45 of 53' 7.0 RISK EVALUA TION l
The first step in the risk evaluation is the defining of the risk segments. Risk segments 1
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 12 risk segments of which 2 are high risk (risk category 2), 3 are medium risk (risk category 4) and 7 are low risk (5 are risk category 6 and 2 are risk category 7). The risk segments are identified in Table 7 below.
(d pd ABB Combustion Engineering Nuclear Operations
ARR D'1 W W Calculation NO. A-PENG-CALC-014, Rev. 00 Page 46 Of 53 Table 7 Risk Segment iden66 Cation Risk SegmentID Consequence 10 Damage GroupID Msk Region P? ping Line Nos.
Msk Segment Start Pbint Msk Segment Emf P6 int Category faHure P6tential Msk Category isometric Drawings CVCS-R 01A-1 CVCS-C01A CVCS-N Medner 2CCA-26-2 *
(2) Downsweem of 2CVC28C (1) lkstroom side of 2* etbew High None 4
(112CCA-26-14 Sh.1 (2) weem ot 2CVC1188 2CCA-26-15 Sh.1 (2) 2CCA-26-16 Sh 1 CVCS R 01A-2 CVCS-CotA CVCS-T High 2CCA-26-2*
(1) Wstream side of 2* ebow (11 CoM Log *C*
- ltem 6 High Smallleek 2
til 2CCA-26-14 Sh.1 CVCS R-018-1 CVCS-C01B CVCS-N Medium 2CCA-27-2*
(2) Downsueem of 2CVC-288 (1) Wserem side of 2* elbow
- from 5 High None 4
(1) 2CCA-27-1 Sh.1 (2) Wsm of 2CVC1186 2CCA-27-2 Sh. I 1212CCA-27-3 Sh.1 CVCS-R 018-2 CVCS-C01B CVCS-T High 2CCA-27-2 *
(1) Wsweem side of 2* elbow fil CoMleg *B*
High Smen Leek 2
(1) 2CCA-27-1 sh. I
~
CVCS-R 02 CVCS-C02 CVCS N Medium 2CCA-12-2*
(1) CoMleg "A*
fil Wstroom of 2RC-4A High None 4
(1) 2CCA-12-1 Sh.1 CVCS-R 03 CVCS-CO3 CVCS N Low 2CCA-12-2
- Ill Downserem of 2RC-4A.
iff Wstream of 2RCSA Medium None 6
(1) 2CCA-12-1 Sh.1 CVCS-R 04A CVCS-004A CVCS-N Low 2CCA-26-2
- Ill Downsurem of 2CV-4831-ill Wstream of 2CVC28C Low None 7
(1) 2CCA-26-16 Sh.1
'T CVCS-R-048 CVCS-C-048 CVCS N Low 2CCA-27-2*
- (1) Downseesm of 2CV-4827-til Wstreem of 2CVC288 low None 7
til 2CCA 27-3 Sh.1 (2) 2CCA-27-4 Sh.1 CVCS-R 05 CVCS-COS CVCS-N Low 2CCA-27-2 *
(1) Downstream of 2CVC-til Wstroom of 2CVC1187 Medium None 6
til 2CCA-27-3 Sh.1 CVCS-R-06 CVCS-C06 CVCS-N Low 2CCA-26-2*
(1) Downstream of 2CVC-(1) Wstream of 2CVC-1189 Medium None 6
(1) 2CCA-26-16 Sh.1 l
ABB Combustion Eng _ ing Nuclear Operations l
nuo Calculation No. A-PENG-CALC-Old, Rev. 00 Page 47 of 53 Table 7 Risk Segrnent identHication (Cont'd)
Msk Segment ID Consepence ID Damage GroupID Msk Region Pipeg Line Nos.
Msk Segment Start P6Ent Msk Segment EndP66nt Category FeMure Potentief Msk Category isometric Drawings CVCS-R 07 CVCS-C Of CVCS-N low 2CCA-12 2*
(11 Doomsveam of 2CV-4820-(2) 2* x 1* Reducmg kwert -
2 trem 35 Madnsn None 6
2CCB 1-4*
(2) 2* x 1* Reducmg kwert -
?CCB-1-2 %
- Item 33 2CCB 1-2*
(2) 2* x %
- Rodksema kwart -
2CCB 2-2% ~
trem 26 2CCB 2-2*
121 Penewetion 2P-14 til 2CCA-12-1 SIL 1 2CCG-1-1 (2) 2CCB-2-1 CVCS-R-08 CVCS-C-08 CVCS N low 2CCB.2-2*
(21 Pmeestion 2fL14 (11 (4wweem of 2CV-4816 Medison None 6
(1) 2CCB 2-2 Sh.1 g,y (212CCB 2-2 Sh. 2 g, 3g (2) 2"x 1*Redkocing kwert -
trem 29 ABB Combustion Engineering Nuclear Operations
A kB
- %IFIF Calculation No. A PENG-CALC 014. Rev. 00 Page 48 of 53 To facilitate application of the sampling percentages to determine the inspection scope, ISIS combines like segments (i.e., same consequence category and damage group) into segment groups. A total of 4 segment groups have been identified and are summarized in Table 8 below.
Table 8 Risk inspection Scope Segment Consequence FeHwe Risk Risk Total Selections Selections Groups Category Potential Repton Category Welds Requked Mode CVCS-00t High None Medium 4
83 9
9 CVCS-002 High SmallLeak High 2
t0 3
4 CVCS 003 Low None Low 7
to 0
0 CVCS 004 Medium None low 6
8t 0
0 i
8.0 ELEMENT SELECTION 7
l The number of elements to be examined as part of t'oe risk-informed developed program l
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 of individualinspection locations within a risk category depends upon the relative
)
severity cf 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 locations (i.e., structural discontinuities) of high stress and/or high fatigue usage. An inspt.ction for cause process shall be implemented utiliting examination methods and volumes defined specifically for the degradation mechanism postulated to be active at the inspection location.
l Tables 9 and 10 depict the element selections ar.d other pertinent information (e.g.,
examination methods and volumes, basis for selection) for risk significant segment groups CVCS-001 and CVCS-002. As required by the Risk Inspection Scope of Table 8, a total of i
13 elements have been selected for examination, including 9 elements from segment group CVCS-001 and 4 elements from segment group CVCS-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.
Currently, no specific guidance is provided in Reference 9.1 regarding appropriate examination methods and volumes for risk category 4 (i.e., no failure potentialidentified) 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 l
fatigue.
r
(
I 1
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g ]. f7
(
- "EFIF Ca!Cu!ation No. A-PENG-CALC-014, Rev. 00 Page 49 Of 53 Table 9 Element Selection - Risk Category 4 Segment Group Consequence FaRure Potential Riek Categeg Rient Region Total S of elemente 10% of elemente CVCS-001 High None Identified 4
Medium 83 9
Bemente Selected Line No.
Exem Method Risk segment D Description leo Deva No.
Exam Vohene Ca.--=,-.ee /DM Groep D's Reeson for Secocean 37-009 2CCA-12-2
- Voksnetric CVCS R-02 h the obsence of anyidentified demoge rmt L.
e Ngher stress (eg.101 element (nodo points 21/22 Calc No. 85-E-0055-16} in the Elbow-to-Pipe Weld 2CCA 12-1 Sh. I figure No. 7.1-1 CVCS C-02 / CVCS-N portion of this risk segment which is potentiaRy subjected to thermal transients (ATis accepteNe, see 4th teNo of Section 5.2.1.21 during hss ofletdown or hss of charging events, has been selected.37-010 2CCA 2
- Vohemetric CVCS R-02 h the obsence of any identified demoge.ad_... the 2nd twghest stress (eq.101 element (node points 25/26, Cole No. 85 E-0065-16)
Pfpe-to Bbow Weld 2CCA 12-1 Sh.1 Rgure No. 7.1-1 CVCS-C-02 / CVCS-N in me WM of mis risk a which is potenddy %'ected to therme! transients (ATis eccepteble, see 4th teNe of Section 5.2.1.21 dtring %ss ofletdown or kss of charging events, has been selected.37-011 2CCA-12-2
- Vohnnetric CVCS-R-02 h the obsence of any identified demoge. meat-;,.ea, the tughest stress (eq.101 element (node points 26/27. Calc No. 85-E-0055-161 Bbow-to-Pipe Weld 2CCA-12-1 Sh.1 Rgure No. 7.1-1 CVCS-C-02 / CVCS-N in the portion of this risk segment which is potentiaRy subjected to thermal transients (ATis eccepteNo, see 4th teNo of Section 5.2.1.21 during hss ofletdown or loss of chergeg events, hos been selected.
40005 2CCA-2 7-2
- Voksnetric CVCS-R 018-1 h the obsence of any idendfied demogo aimt ;~.au, the Nghest stress (eq.101 eternent (node point 110. Calc No. 85-E-0055-031 k Valve-to-Pipe Wald 2CCA-27 3 Sh.1 Rgure No. 7.1-1 CVCS C-01B / CVCS-N this risk segment, wNch is potentia #y sub ected to thermet transients t
(AT is accepteNo, see 2nd raNe of Section 5.2.1.21 dunng kss of charging or hss of letdown events, and which has also been cut out once (FW5C11 and rewelded (creating Ngher residuet stressest, hos been selected.
40003 2CCA-27 2*
Volumetric CVCS-R 018-1 h the obsence of any identified damage enechanisms, the 3-d Nghest stress (eq.101 element (nodo points 10W191, Cole No. 85-EM5-031 Pipe-to-Dbow Weld 2CCA-27-3 Sh.1 Rgure No. 7.1-1 CVCS C-018 / CVCS-N k this risk segment wNch is prontidy s&ted to W transients (ATis eccepteNo, see 2nd teNe of Section 5.2.1.2} during hss of charging or kss ofletdown events, has been selected.
ABB Combustion Engi_
ing Nuclear Operations
q n
A A t l [,t 7%FBF Calcu!ation NO. A-PEN'l-CALC-014, Rev. 00 Page 50 of 53 Table 9 Element Selection - Risk Category 4 (Cont'd)
Somment Group Ceneequence FeRure Potential Risk Category Riek Re&
Total 9 et elemente 10% of e umerne n
CVCS 001 High None kfentified 4
Medium 83 9
Barnente Selected Line No.
Exam Method Risk Segment D Descript{on leo Dwg No.
Exam Vohene Consequence / DM Grosp D'e Reenen for Sedeceen 40009 2CCA 27-2*
Vokenetric CVCS-R 018-1 h the obsence of any identified demoge i mf-.i-.u the 2nd twghest stress (eg.101 element (nodo points 99/100. Cole No. 85-E-0055-031 Dbow-to-Mpe Weld 2CCA-27-3 Sh.1 Figure No. 7.1-1 CVCS C 018 / CVCS-N in tNs risk segment, which is potentieRy sub ected to thermal t
transients (AT is accepteNo, see 2nd teNe of Section 5.2.1.21 during kss of charging or kss ofletdown events, and which has also been cut out once and repeired twice (Fu9CIR2: and rewekied (creetM Ngher residuelstresses), has been selected.41-003 2CCA 26-2*
Vokmetric CVCS R 01A-1 k the obsence of any identified demege
.mf-.i.., the twghest stress leg.101 element (nodo point 140 Cole No. 85 E-0055-03) in Velve-to-hpe Weld 2CCA-26-16 Sh.1 Rgure No. 7.1-1 CVCS C-01A / CVCS-N tNs risk segment, which is also potentiocy subjected to thermet transients (AT is accepteNo, see 2nd teNo of Section 5.2.1.21 durirog kss of charging or kss ofletdown events, has been selected.
41-003C 2CCA-26-2
- Vokanetric CVCS-R-01A-1 h the absence of any identified damage i-mf- --.e. the 2nd Nghest stress (eq.101 element (node pok:t 278, Cole No. 85-E-0055-031 in Tee-to-Mpe Weld 2CCA-26-16 Sh.1 Figure No. 7.1-1 CVCS-C-01A / CVCS-N tNs risk segment, which as also potentiety subjected to therme!
transients (ATis eccepteNo, see 2nd teNo of Section 5.2.i.21 during kss of charging or kss ofletdown events, has been selected.41-008 2CCA-26-2~
Voksnetric CVCSR 01A 1 h the obsence of any identified demoge mecherwsms, a higher strees
% 101 element (node points 115/116, Calc No. 85 E-0055 031 k Pl(pe-to-Obow Weld 2CCA-26-16 Sh.1 Figure No. 7.1-1 CVCS-C-01A / CVCS-N this risk segment, which is also potentiaMy subjected to thermet transients (ATis necepteNo, s m 2nd teNo of Section 5.2.1.21 during kss of charging or kss oflordown events, hos been selected.
ABB Combustion Engineering Nuclear Operations
I ARn 7% FIF Calculation No. A-PENG-CALC-014, Rev. 00 i
Page 51 ' f 53 O
Table 10 Eternent Selection - Risk Category 2 Seement Gmap Coneentuence l'eRure Potentief Risk Category Risk Rea on Tata 8 et elemente 25% of elemente s
s CVCS-002 High Smen Leek 2
High 10 3 spicaed 41 l
Bemerne Selected Line No.
Esem Method Risk Segment D Deecripden leo Dwg No.
Euem Vdlume Consequence /DM Groep D's Reeson hre Selecten 40 025 2CCA-27-2
- Vokanetric CVCS R D18-2 TNs risk segment is potentioWy s@ected to thermal transients durmg kss of chargeg or kss ofletdown events (see 1st teNe of Secnon Elbow-to Safe End Wald 2CCA-27-1 Sh.1 Egure No. 7.1-1 CVCS-C 018 / CVCS-T 5.2. 1.21.
TNs element. wNch has also been cut out once inVICI and rewelded (creating Ngher residual stresses), has been selected since it is subjected to the 2nd Nghest cumulative fatigue usage in tNs risk segment 40 026 2CCA-27-2
- Voksnetric CVCS-R 018-2 TNs risk segment is potendeMy sukeeted to therme! transients dwing kss of charging or loss of letdown events (see 1st teNe of Secten Safe End to-Nozz!e Weld 2CCA-27-1 Sh.1 Rgure No. 7.1-1 CVCS C 018/ CVCS-T 5.2.1.2).
TNs terminal end element hos been selected shee it is subjected to the tughest cumuledve fatigue usage h tNs risk segment.41-039 2CCA-26-2
- Vokanetric CVCSR 01A-2 TNs risk segment is potentiecy subjected to thermet transients during kss of charging or kss ofletdown events (see 1st teNe of Section L%ow-to-Safe End Wald 2CCA-26-14 Sh.1 Rgure No. 7.1-1 CVCS C-01A / CVCS-T 5.2.1.2).
TNs element has been selected since it is subjected to the 2nd Nghest cur.,ufe6ve fatigue usage in tNs risk segment.41-040 2CCA-26-2*
Vokanetric CVCSR-01A-2 TNs risk segment is potentioWy subjected to thermet transients d6 ring kss of charghg or kss of letdown events (see 1st teNe of Section Sete End-to-Nozzle Weld 2CC4-26-14 Sh.1 Rgure N&. 7.1-1 CVCS C 01A / CVCS-T 5.2.1.21.
TNs termenal and element has been selected since it la subjected to the lughest cumulative fatigue usage h tNs risk segment.
t ABB Combustion Eng. _ -ing Nuclear Operations
A Ik It A"EIF BF 3
Calculation No. A PENG CALC-014, Rev. 00 (O
Page 52 of 53
- 9. 0 REFERENCES 9.1
' Risk informed Inservice Inspection Evaluation Procedure,' EPRI Report No. TR-106706, Interim Report, June 1996.
- 9. 2 EPRIInservice inspection Software (ISIS ~),1996.
9.3 Arkansas Nuclear One Unit 2,
- Safety Analysis Report,' Amendment No.13.
- 9. 4
" Design Specification for ASME Section ill Nuclear Piping for Arkansas Nuclear One Unit 2, Arkansas Power and Ught Company,' Specification No. 6600-M-2200, rievision 9.
- 9. 5
'ANO-2 Str.tS Components Database,"(Plant Piping Une List (M 2083), dated 3 31 96).
9.6 "ANO-2 ISI Plant Piping Une Ust," from Revision 4 of ANO-2 Inservice inspection Plan.
- 9. 7
" Instruction Manual, Reactor Coolant Pipe and Fittings," Arkansas Nuclear One Unit No. 2 C.E. Book No. 73470, June 1974.
1 9.8
' Summary of Combustion Engineering Owners Group Programs in Response to NRC Bulletin 88-08,' CEOG Task 866, Report No. CE NPSD-1043, Revision 00, v
September 1996.
9.9
' Technical Specification for Insulation for Arkansas Nuclear One Unit 2 of the Arkansas Power and Ught Company," Specification No. 6600-M-2136, Revision 9.
9.10
" Primary Chemistry Monitoring Program," Procedure No. 1000.106, Revision 4, 9.11 "EPRI Fatigue Management Handbook," Report No. TR-104534-V1,-V2,-V3, V4, Project 3321-01, Final Report, December 1994.
9.12
' Pipe Cracking in PWRs with Low Pressure Borated Water Systems,' EPRI Report No. IVD-3320.
9.13
" Flow Accelerated Corrosion Prevention Program,' HES-05, Revision 1.
9.14 Arkansas Nuclear One, Unit 2, "TechnicalSpecifications, Appendix A to Ucense No.
NPF-6, Amendments Nos.173 and 174."
9.15 Gaertner, J.
P., et. al., " Arkansas Nuclear One Unit 2 Intemal Flood Screening Study," prepared for Entergy Operations, Inc., Calculation No. 89-E-0048-35, Rev.
O, May 1992.
9.16
" Arkansas Nuclear One Unit 2 Probabilistic Risk Assessment, Individual Plant Examination Submittal," 94-R-2005-01, Rev. O, August 1992.
ABB Combustion Engineering Nuclear Operations
N !I !I.
MIfIF Calculation No. A.PENG CALC-014, Rev. 00 Page 53 of 53' 9.17 Entergy, Arkansas huclear One Unit 2, Drawings:
1.0 Orawing No. M 2230, Sheet 1, Rev. 71, Sheet 2, Rev. 26;
- Piping &
Instrumentatic n Diagram Reactor Coolant System'.
- 2. 0 Drawing No. I4 2231, Sheet 1, Rev.121. Sheet 2, Rev. 68, Sheet 3, Rev.
5; ' Piping & h ostrumentation Diagram Chemical & Volume Control System".
3.0 Drawing No. 2CCA 121, Sheet 1, Rev.11, Sheet 2, Rev. 5; "Large Pipe Isometric Reactor Coolant Letdown to Control Valve 2CV-4821-l'.
- 4. 0 Drawing No. 2CCA 2614, Sheet 1, Rev.16; 'Small Pipe Isometric Chemical
& Volume Control'.
- 5. 0 Drawing No. 2CCA 2616, Sheet 1, Rev. 7; 'Small Pipe isometric Chemical
]
& Volume Control".
- 6. 0 Drawing No. 2CCA 26-16, Sheet 1, Rev.11; "Small Pipe isometric Chemical
& Volume Control'.
7.0 Drawing No. 2CCA-271, Sheet 1, Rev.19; "Small Pipe isometric Charging inlet from Regenerative Heat Exchanger".
. 9. 0 Drawing No. 2CCA 27 2, Sheet 1, Rev. 8; "Small Pipe Isometric Charging inlet from Regenerative Heat Exchanger to 2CCA 6".
- 9. 0 Drawing No. 2CCA-27 3, Sheet 1, Rev.10; "Small Pipe isometric Charging inlet from Regenerative Heat Exchanger to 2CCA-27 3".
10.0 Drawing No. 2CCA 27 4, Sheet 1, Rev.13; 'Small Pipe isometric Charging inlet from Regenerative Heat Exchanger".
11.0 Drawing No. 2CCB 1 1, Rev. 9; "Large Pipe isometric Letdown from 2CV.
48212 to Regenerative Heat Exchanger 2E-23'.
12.0 Drawing No. 2CCB-21, Rev.10; "Small Pipe isometric from Regenerativa Heat Exchanger 2E-23 to Letdown Heat Exchanger 2E-29 and Reactor Drain Tank 2T 68".
13.0 Drawing No. 2CCB-2 2, Sheet 1, Rev.14; "Small Pipe Isometric Chemical &
Volume Control from Regenerative Heat Exchanger to Letdown Heat Exchanger".
14.0 Drawing No. 2CCB 2 2, Sheet 2, Rev. 6; *Small Pipe isometric Chemical &
Volume Control from Regenerative Heat Exchanger to Letdown Heat Exchanger".
9.18
' Enhanced Fatigue Manage nent Handbook Screening Criteria," SIC 97 002, March, 1997 9.19
" Enhanced Screening Procedure for Thermal Transients," SIC-97 005, March 1997.
9.20 Interoffice Correspondence from A. V. Bauer to Quality Records, Letter No. PENG-97 140, " Submittal of SIA Calculations," dated July 21,1997.
O 1
ABB Combustion Engineering Nuclear Operations
_.___-__...m.__
Calculation No. A PENG CALC Old, Rev, 00 Page At of A22
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i i
l 1
i i
J l
APPENOlX A
'FMECA - CONSEQUENCE INFORMA TION REPORT" (Attachment Pages A1 A22)
'O ABB Combustion Engineering Nuclear Operations
FMECA - Consequence Information Report Calculadon No. A PENG-CALC-04. Rev. 00 14 ser-97 Page A2 of A22 Consequence ID: CVCS-C 01 A l
Consequence Descdption: Loss of reactor coolant sia the charging line to the 2P32C cold leg occurs during l
normal power operation.
';reak Size:
Large isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation, and in the piping from downstream of charging line check valve 2CVC-28C to the 2P32C cold leg. This consequence evaluation includes the welds in the applicable portions ofline 2CCA 26-2" A failure in this segment would result in a small Loss of Coolant Accident (LOCA) This is characterized by a decrease in RCS pressure, followed by automatic plant shutdown in order to bnng the plant to a stable state. The lost coolant drains to the containment sump as expected and is then recirculated by the HPSI pumps. The failed segment cannot be isolated because there are no isolation valves in the above line segment.
Spatial Effects: Containment Effected Location: Containment Building Spatial Effects Comments: A dynamic analysis (SAR Section 3.6.4.2.9.2) which included the above line has been performed. The analysis concluded that systems required to mitigate the consequence of the break will not be impaired byjet impingement or uncontrolled w hipping of this line. Restraints in addition to existing piping and structures will protect the shutdown cooling valves and ensure that shutdown cooling is available if required.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units, SITS, and their electrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (S AR Section 6.2.2.2.2). In addition, all safety injection components located inside the containment and their associated electrical equipment have been designed to withstand the LOCA emironment (SAR Section 6.3.2.12.1). Hence, the impact of spatial effects due to a failed segment is assumed to be negligible.
Initiating Event: I initiating Event ID: S Initiating Event Recovery: Based on the ANO-2 IPE (Report 94 R 2005 01, Rev. 0), the following equipment is required to mitigate a small LOCA.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
(b) For RCS inventory control (i.e., injection mode), one of three HPSI pumps.
(c) For long term RCS inventory control and heat removal (i. c., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shutdown cooling heat exchanger or one of two CS pumps and two containment cooling units.
If a loss of feedwater occurs, core heat removal can be accomplished by "once through cooling" using the LTOP or ECCS vent vrsives. Automatic actuation of the
[
reactor protection system and the engineered safety features actuation system occurs l
A FMECA - Consequence Information Report Catalatwo Na A PENG C4f-014. Rev. 00 0
14-s*r-91 Page A3 of A22
= _, _ _.
in responic to this initiating event.
Loss of System: S System IPE ID:
Charging System Recova.g; It:.ccovery ticlerging is not expected during the response to the LOCA. The HPSI system can provide the needd makeup and boration required for mitigating a small LOCA. All engineered safey featu es required to mitigate this initiating event are automatically actuated.
Las of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: A small Loss of Coolant Accident (LOCA) occurs due to the failed segment. Because the segment failure causes an initiating event (i c., design basis category IV event), and based on Table I which was developed specifically for ANO-2 using the guidelines provided in Tables 3.1 end 3.4 of the EPRI procedure (EPRI TR 106706), a IDGH consequence category is assigned.
Consequence Category: HIGH O
Con.equence aank O
O d
FMECA - Consequence information Report Cakulation No. A-PENG-C4LC-OH, Rn 00 14ser-91 Page A4 of A22 Consequence ID: CVCS-C OlB Consequence
Description:
Loss of reactor coolant via the charging line to the 2P32B cold leg occurs during normal power operation.
Break Stre:
Large Isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation, and in the piping from downsteam of charging line check valve 2CVC-28B to the 2P32B cold leg. This consequence evaluation includes the welds in the applicable portions of line 2CCA 27 2",
A failure in this segment would result in a small Loss of Coolant Accident (LOCA). This is characterized by a decrease in RCS pressure, followed by automatic plant shutdown in order to bring the plant to a stable state. The lost coolant drains to the containnent sump as expected and is then recirculated by the HPSI pumps. The failed segment cannot be isolated because there are no isolation valves in the above line segment.
Spatial Effects: Containment Effected Location: Containment Building Spatial Effeeta Comments: A dynamic analysis (SAR Section 3.6.4.2.9.2) which included the above line has been performed. The analysis concluded that systems required to mitigate the consequence of the break will not be impaired by jet impingement or uncontrolled whipping of this line Restraints in addition to existing piping and structures will protect the shutdown cooling valves and ensure that shutdown cooling is available if required.
C.;rtain safety related components v hich are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units SITS, and their electrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (SAR Section 6.2.2.2.2). In addition, all safety injection components located inside the containment and their associated electncal equipment have been designed to withstand the LOCA emironment 'SAR Section 6.3.2.12.1). Hence, the impact of spatial effects due to a failed segment is assumed to be negligible.
Initiating Esent: 1 initiating Event ID: S Initiating Event Recosery: Based on the ANO-2 IPE (Re:
94 R-2005-01, Rev. 0), the following equipment is required to mitigate a small L(A.A.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
(b) For RCS inventory control (i.e., injection mode), one of three HPSI pumps.
(c) For long term RCS inventory control and heat removal (i. c., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shutdown cooling heat exchanger or one of two CS pumps and two containment cooling units, if a loss of feedwater occurs, core heat removal can be accomplished by "once through cooling" using the LTOP or ECCS vent valves. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs
FMECA - Consequence Information Report Calculatwn No. A PENG CALC-Old, Rev 00 14-Sep 91 Page A3 cf A22 in response to this initiating event.
Loss of System: S System IPE ID:
Charging System Recovery: The recovery of charging is not expected during the response to the LOCA. The HPSI system can provide the needed makeup and boration required for mitigating a small LOCA. All enginected safety features required to mitigate this initiating event are automatically actuated.
Loss of Train: N Train ID:
N/A Train Recoscry: N/A J
Consequence Comment: A small. 5s of Coolant Accident (LOCA) occurs due to the failed segment. Because the segms i failure causes an initiating event (i.e., design basis category IV event), and u
based on Tai le I w hich was developed specifically for ANO 2 using the guidelines provided ir. 's tbles 3.1 and 3.4 of the EPRI procedure (EPRI TR 106706), a HIGH consequence category is assigned.
Consequence Category: HIGH O
Consequence Rank O
O l
1 O
FMECA - Consequenet Information Report Calculanon No. A.PENG.C4LC-Old, Rev 00 la ser 91 Page A6 of A22 Consequence ID: CVCS-C-02 Consequence
Description:
less of reactor coolant via letdown line occurs during normal pc a operation.
Break Slic:
Large Isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation, and in the piping from RCS cold leg 2P32 A to upstream ofletdown isolation valve 2CV-4820 2. This consequence evaluation includes the welds in the applicable portion ofline 2CCA 12 2",
A limiting failure in the segment would result in a small Loss of Coolant Accident (LOCA).
This is characterized by a decrease in RCS pressure, followed by automatic plant shutdown in order to bring the plant to a stable state. The lost coolant drains to the containment sump as expected and is then recirculated by the HPSI pumps. For the limiting segment failure, isolation cannot be accomplished because there are no valves in the line segment of concern.
Spatial Effects: Containment Effected Location: Contaitunent Building Spatial Effects Comments: A dynamic analysis (SAR Section 3.6.4.2.10.2) which included the above lines has been performed. The analysis concluded that systems required to mitigate the consequence of the break will not be impaired byjet impingement or uncontrolled whipping of this line. Restraints in addition to existing piping and structures will protect the systems required for safety from being affected by the segmer.t failure.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units, SITS, and their electrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (S AR Section 6.2.2.2.2). In addition, all safe,y injection components located inside the containment and their associated electrical equipment have been designed to withstand the LOCA emironment (SAR Section 6.3.2.12.1). Hence, the impact of spatial efTects due to a failed segment is assumed to be negligible.
Initiating Event: I initiating Event ID: S Initiating Event Recovery: Based on the ANO-2 IPE (Report 94-R-2005-01, Rev. 0), the following equipmer. is required to mitigate a small LOCA.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
(b) For RCS inventory control (i.e., injection mode), one of three HPSI pumps.
(c) For long term RCS inventory control and heat removal (i. c., recirculation mode), one of three FPSI pumps and one of two CS pumps with an associated shutdown cooling heat exchanger or one of two CS pumps and two containment cooling units.
If a loss of feedwater occurs, core heat removal can be accomplished by "once through cooling" using the LTOP or ECCS vent valves. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiating event.
FMECA - Consequence Information Report C**laai"' No A PENG. Calf Old, Rev. 00
\\
144*p-91 Page At of A22 less of System: S System IPE ID:
Letdown System Recovery: Attempts to recover letdown flow are not expected because this portion of the CVCS is not needed for mitigating a small LOCA. All enginected safety features required for mitigating tids initiating event are automatically actuated.
Loss of Train: N Train ID:
N/A Train Recovery: N/A Consenynce Comment: A small Loss of Coolant Accident (LOCA) occurs due to the failed segment. Because the segment failure causes an initiating event (i.e., design basis category IV event), and based on Table I which was developed specifically for ANO 2 using the guidelines provided in Tables 3.1 and 3.4 of the EPRI procedure (EPRI *1%106706), a HIGH consequence category is assigned.
Consequence Categon: HIGH O
Consequence an k O
O r\\
U
Calculation No. A PENG. CALC-014. Rn 00 FMECA - Consequence Information Report t 4.seys7 Page A8 of A22 Consequence ID: CVCS-C-03 Consequence
Description:
Potentially loss of coolant occurs sia RCS cold leg 2P32 A drain line.
Break Slic:
Large Isolability of Break: No ISO Comments: ne break is postulated to occur during plant shutdown, and in the piping between RCS cold leg drain line manual valves 2RC 4 A & 2RC 5 A. The piping is isolated during normal power operation. If drain valve 2RC-4 A remains closed, a segment failure would not cause an initiating event. The drain line is also not needed to support or accomplish any of the safety functions required for mitigating a design basis event. The resulting corsequence would be less significant than the consequence of a small LOCA (see Section 4.1.6) which challenges mitigating systems. Therefore, only the case for a potential small LOCA is described herein.
This consequence evaluation includes the welds in the applicable portion ofline 2CCA 12 2",
Failure of drain valve 2RC-4 A followed by a segment failure would result in a small LOCA.
This is characterized by a rapid decrease in RCS pressure, followed by automatic plant shutdown in order to bring the plant to a stable state. De lost RCS inventory would drain to the containment sump as expected and would then be recirculated by the HPSI pmps. The failed segment cannot be isolated because there are no isolation valves upstream of valve 2RC-4 A.
Spatial Effects: Containment Effected Location: Containment Building Spatial Effects Comments: A dynamic analysis of the RCS (S AR Section 3.6.4.2.1) lines has been performed.
The analysis concluded that systems required to mitigate the consequence of the break will not be impaired byjet impingement or uncontrolled whip of this drain line.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units, SITS, and their electrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (SAR Section 6.2.2.2.2). In addition, all safety injection components located inside the containment and their associated electrical equipment have been designed to withstand the LOCA emironment (SAR Section 6.3.2.12.1). Hence, the impact of spatial effects due to the segment failure is assumed to be negligible.
Initiating Esent: 1 Initiating Event ID: S Initiating Event Recovery: Based on the ANO-2 IPE (Report 94-R 2005-01, Rev. 0), the following equipment is required to mitigate a small LOCA.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
(b) For RCS inventory control (i.e., injection mode), one of three HPS! pumps.
(c) For long term RCS inventory control and heat removal (i. c., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shutdown cooling heat exchanger or one of two CS pumps and two containment cooling units.
FMECA - Consequence Information Report Calculation No. A.PENG CALC 414, Rev. 00 l4 sw91 Page A9 of A22 If a loss of feedwater occurs, core heat removal can be accomplistd by "once through cooling" using the LTOP or ECCS vent valves. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiating event.
14ss of System: N System IPE ID:
N/A System Recovery: N/A IAss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: Failure of drain valve 2RC-4 A to remain closed will expose the segment to the operating temperatures and pressures of the RCS. By exposing the segment to operating conditions, the potential for a small LOCA exists. The combined effect of a passive failure of the manual valve and the conditional core damage probability for a small LOCA (Table 1) results in a MEDIUM consequence (see Section 4.1.6).
Consequence Category: MEDIUM O
Consequence Rank D
cJ
e FMECA - Consequence Information Report Calculation No A PENG-Calf 014. Rev. 00 l<.sep 91 Page A!0 of A22 Consequence ID: CVCS-C 04 A Consequence
Description:
Loss of charging flow to 2P32C cold leg occurs due to a break in the charging line inside containment.
Break Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation. The piping from downstream of either charging line isolating valve 2CV-4831 2 to upstream of charging lin.; ; heck valve 2CVC 28C is included. Failure of the segment during power operation followed by failure of the check vaht 2CVC 28C to close will result in a small LOCA. The consequence of this scenario is more risk significant than the loss of charging to the RCS. Therefore, only the case of a potential small LOCA is described herein. This consequence evaluation includes the welds in the applicable portions ofline 2CCA 26 2".
The potential failure and the resulting small LC,lA would be characterized by a rapid decrease in RCS pressure, followed by automatic plant shutdown in order to bring the plant to a stable state. The RCS inventory would drain to the containment sump as expected and would then be recirculated by the HPSI pumps. The failed segment cannot be isolated because there are no isolation valves downstream of the check vahr.
Spatial Effects: Containment Effected location: Containment Building Spatial Effects Comments: A dynamic analysis (SAR Section 3.6.4.2.9.2) which included the above line has been performed. The analysis concluded that systems required to mitigate the consequence of the break will not be impaired by jet impingement or uncontrolled whipping of this line. Restraints in addition to exiaing piping and structures will protect the shutdown cooling valves and ensure that shutdown cooling is available if required.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units, SITS, and their electrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (S AR Section 6.2.2.2.2), in addition, all safety injection components located inside the containment and their associated electrical eqmpment have been designed to withstand the LOCA emironment (SR Section 6.3.2.12.1). Hence, the impact of spatial effects due to a failed segment is assumed to be negligible.
Initiating Event: 1 Initiating Event ID: S Initiating Event Recovery: Ba. sed on the ANO-2 IPE (Report 94 R 2005-01, Rev 0), the following equipment is required to mitigate a small LOCA.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
(b) For RCS inventory control (i.e., injection mode), one of three HPSI pumps.
(c) For long term RCS imentory control and heat removal (i. c., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shutdown cooling heat exchanger or one of two CS pumps and two containment cooling units.
-~_--.
A FMECA - Consequence Information Report Calculatum h A PENG CALC-Oldev. 00 0
I*59" Page All of A22 If a loss of feedwater occurs, core heat removal can be accomplistd by "once through cooling" using the LTOP or ECCS vent vahes. Automatic actuation of the reactor pittection system and the enginected safety features actuation system occurs in response to this initiating event.
Loss of System: N System IPE ID:
N/A System Recovery: N/A Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: A segment failure followed by failure of the charging line check valve to close would cause a small LOCA. The combined effect of a check valve failure and the conditional core damage probability for a small LOCA (Table 1) results in a LOW consequence (see Sect:0n4.1.6).
Consequence Category: low O
Consequence aank O
O 1
O
FMECA - Consequence Information Report Calculatwo No A-PENG CNf-Old, Rev. 00 t 4-Scr91 Page Al2 of A22 Consequence ID: CVCS-C h4B Consequence
Description:
Loss of charging flow to the 2P32B cold leg occurs due to a break in the charging line inside containment.
l Break Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation. The piping frore downstream of either charging line isolating valve 2CV-4827-2 to upstream of charging line check vahr 2CVC 28B is included. The line segment downstream of check valve 2CVC 26 is also included. Failure of the segment during power operation followed by failure of the check vaht 2CVC 288 to close will result in a small LOCA. The consequence of this scenario is more risk significant than the loss of charging to the RCS. Therefore, only the case of a potential small LOCA is described herein. This consequence evaluation includes the welds in the applicable portions ofline 2CCA 27 2" The potential failure and the resulting small LOCA would be characterized by a rapid decrease in RCS pressure, followed by automatic plant shutdown in order tc oring the plant to a stable state. The RCS inventory would drain to the containment sump as expected and would then be recirculated by the HPSI pumps. The failed segment cannot be isolated because there are no isolation valves downstream of the check vahr.
Spatial Effects: Containment Effected Location: Containment Building Spatial Effects Comments: A dynamic analysis (SAR Section 3.6.4.2.9.2) which included the above lines has been performed. The analysis concluded that systems required to mitigate the consequence of the break will not be impaired byjet impingement or uncontrolled whipping of this line. Restraints in addition to existing piping and structures will protect the shutdown cooling valves and ensure that shutdown cooling is available if required.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units, SITS, and their electrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (S AR Section 6.2.2.2.2). In addition, all safety injection components located inside the containment and their associated electrical equipment have been designed to withstand the LOCA emironment (SR Section 6.3.2.12.1). Hence, the impact of spatial effects due to a failed segment is assumed to be negligible.
Initiating Event: 1 Initiating Event ID: S Initiating Event Recovery: Based on the ANO-2 IPE (Report 94-R 2005-01, Rev. 0), the following equipment is required to mitigate a small LOCA.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
(b) For RCS imentory control (i.e., injection mode), one of three HPSI pc nos.
(c) For long term RCS inventory control and heat removal (i. e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shutdown cooling heat exchanger or one of two CS pumps and two containment
FMECA - Consequence Information Report Cablanon No A PEAri-CALC 014. Rev. 00 14-Sep97 Page A13 of A22 cooling units.
If a loss of feeduler occurs, core heat removal can be accomplished by "once through cooling" using the LTOP or ECCS vent valves. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiating event.
less of System: N System IPE ID:
N/A System Reemery: N/A Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: A segment failure followed by failure of the charging line check valve to close would cause a small LOCA. The combined effect of a check valve failure and the comiitional core damage probability for a small LOCA (Table 1) results in a LOW consequence (see Section 4.1.6).
Consequence Category: low O
Consequence aank O
V
\\
\\
f O
FMECA - Consequence Information Report Calculatma Na A PENG C4LC-014. Rev 00 s s-Sm91 Page A14 of A22 Consequence ID: CVCS-C 05 Consequence
Description:
Potential loss of coolant occurs via chargmg line (to RCS cold leg 2P32B) drain piping.
Hrcak Size:
Large Isolability of Break: No ISO Comments: The break is postulated to occur during plant shutdown, and in the piping between charging line manual drain valves 2CVC 1186 & 2CVC 1187. The piping is isolated during normal power operation. If drain valve 2CVC 1186 remains closed, a segment failure would not cause an initiating event. The drain line is also not needed to support or accomplish any of the safety functions required for mitigating a design basis event. The resulting consequence would be less significant than the consequence of a small LOCA (see Section 4.1.6) which challenges mitigating systems. Therefore, only the case for a potential small LOCA is described herein.
This consequence evaluation includes the welds in the applicable portion ofline 2CCA 27 2".
Failure of drain valve 2CVC 1186 followed by a segment failure would result in a small LOCA. This is characterized by a rapid decrease in RCS pressure, followed by automatic plant shutdown in order to bring the plant ot a stable state. The loss of RCS inventory would dmin to the containment sump as expected and would then be recirculated by the HPSI pmps. The failed segment cannot be isolated because there are no isolation vahts upstream of valve 2CVC-1186.
Spatial Effects: Containment Effected Location: Containment Building Spatial Effects Comments: A dynamic analysis of the RCS (SAR Section 3.6.4.2.1) lines has been performed.
The analysis concluded that systems required to mitigate the consequence of the break will not be impaired byjet impingement or uncontrolled whip of this drain line.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units, SITS, and their electrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (S AR Section 6.2.2.2.2). In addition, all safety injection components located inside the containment and their associated electrical equipment have been designed to withstand the LOCA emironment (SAR Section 6.3.2.12.1). Hence, the impact of spatial effects due to the segment failure is assumed to be negligible.
Initiating ' Event: 1 Initiating Event ID: S Initiating Event Recovery: Based on the ANO-2 IPE (Report 94 R 2005-01, Rev. 0), the following equipment is required to mitigate a small LOCA.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
(b) For RCS inventory control (i.e., injection mode), one of three HPSI pumps.
(c) For long term RCS imtntory control and heat removal (i. e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shutdown cooling heat exchanger or one of two CS pumps and two containment cooling units.
. - -. - ~
~
FMECA - Consequence information Report Calculation No. A PENGGM-014, Rev. 00 14-ser 91 Page A!3 of A22 If a loss of feeduler occurs, core heat removal can be accomplished by "once i
through cooling" using the LTOP or ECCS vent valves. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs
)
in response to this initiating event, i
1 Loss of Systein: N System IPE ID:
N/A System Recovery: N/A Loss of Train: N Train ID:
N/A Train Recovery: N/A Consequence Comment: Failure of drain valve 2CVC Il86 to remain closed will expose the segment to the operating temperatures and pressures of the RCS. By exposing the segment to operating conditions, the potential for a small LOCA exists. The combined effect of a passive failure of the manual valve and the conditional core damage probability for a small LOCA (Table 1) results in a MEDIUM consequence (see Section 4.1.6).
Consequence Category: MEDIUM O
Consequence Rank O
k V
t 1
i s
l 1
O
FMECA Consequence Information Report Cahla**'n A rma c4w.014.Rn oo I4.s W91 Pogs A16 of A22 Consequence ID: CVCS C 06 Consequence
Description:
Potential loss of coolant occurs via charging line (to RCS cold leg 2P32C) drain piping.
lireak Sire Large Isolability et Break: No 150 Comments: The break is postulated to occur during plant shutdown, and in the piping between charging l
line manual drain valves 2CVC.I188 & 2CVC i189. The niping is isolated during normal f
power operatku if drain vahc 2CVC ll88 remains closed, a segment failure would not cause an initiating event. The drain line is also not needed to su;> port or accomplish any of the safety functions required for mitigating a design basis event. The resulting consequeno: would be less significant th.m the consequence of a small LOCA (see Section 4.1.6) w hich challenges mitigating systems. Therefore, only the case for potential small LOCA is described herein.
This consequence evaluation includes the welds in the applicable portion of line 2CCA 26 2*.
Failure of drain valve 2CVC.1188 followed by a segment failure would result in a small LOCA. This is characterited by a rapid decrease in RCS pressure, followed 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 IIPSI pmps. The failed segment cannot be isolated because there are no isolation valves upstream of valve 2CVC.
I188.
Spatial Effects: Containment Effected Location: Containment Building Spatial 2ffects Comments: A dynamic analysis of the RCS (SAR Section 3.6 4.2.1) lines has been perfonned.
The analysis concluded that systems required to mitigate the consequence of the break will not be impaired by jet impingement or uncontrolled whip of this drain line.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the containment cooling units, SITS, and their electrical equipment. 'Ihe containment cooling units are designed to maintain their functional ir.egrity following a LOCA (S AR Section 6.2.2.2.2). In addition, all safety injection wmponents located inside the contain.nent and their associated electrical equipment hrxe been designed to withstand the LOCA emironment (SAR Section 6.3.2.12.1). llence, the impact of spatial effects due to the segment failure is assumed to be negligible.
Initiating Esent: 1 inillating Event ID: S Initiating Esent Reco cry: Based on the ANO 2 IPE (Report 94.R 2005 01, Rev 0), the following equipment is required to mitigate a small LOCA.
(a) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.
'(b) For RCS inventory control (i.e., injection mode), one of three llPSI pumps.
(c) For long term RCS inventory control and heat removal (i. c., recirculation mode), one of three IIPSI pumps and one of two CS pumps with an anociated shutdown cooling heat exchanger or one of two CS pumps and two containment cool!rg units.
O FMECA - Consequence Information Report Calala*(*
- A FMG44LC OH R'x 00 is-s 91 Page A11 of A22 w
If a toss of feedwater occurs, core heat removal can be accomplished by 'once through cooling" using the LTOP or ECCS vent valves. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiating event.
Less of System: N System IPE ID:
N/A System Recmery: N/A less of Train: N Train ID:
N/A Train Recovery: N/A Conwquence Comment: Failure of drain valve 2CVC ll88 to remain closed will expose the segment to the operating temperatures and pressures of the RCS. By exposing the ugment to operating conditions, the potential for a small LOCA exists. The combined effect of a passive failure of the manual valve and th: conditional core damage probability for a small LOCA (Table 1) results in a MEDIUM consequence (see Section 4.1.6),
Conwquence Category: MEDIUM O
Conwg.c.ce n.nk O
O l
O 6
FMECA - Consequent
- Information Report Cahla'*'h A FMG C4LC OlOrr 00 la ser91 Page AIS of A22 Consequence ID: CVCS-C 07 Consequence
Description:
LOCA occurs due to a letdown line break during normal power operation.
Break Stre:
Large Isolability of Breakt Yes ISO Commenis: The break is postulated to occur during normal power operation, and in the piping from downstream of the letdnwn ime isoladon valve 2CV-4820 2 to containment penetration 2P 14.
This consequence evaluation includes the welds in line 2CCD l 2" and the applicable portions oflines 2CCA 12 2" and 2CCD 2 2".
A limidng failure in the segment would cause a rapid decrease in RCS pressure, subsequent reactor trip and generation of Safety injection Actuation Signal (SIAS). The letdown isolation valve,2CV-4820 2, and the regenerative heat exchanger inlet valve,2CV 4821 1, would close automatically upon receipt of SIAS. llence for this segment failure, the break would be automatically isolated. Because the break flow would initially be sufnciently large to exceed the normal makeup capacity of 144 gpm, this break would be considered a LOCA. The reactor coolant from the break would drain to the containment sump, ne necessary reactor coolant makeup can be provided by the llPSI system once RCS pressure decreases to below IIPSI pump shutofT head following isolation by SIAS. If necessary, RCS pressure relief can be manually initiated.
Spatial Effects: Containment Effected Location: Containment Building Spatial Effects Comments: A dpamic analysis (SAR Section 3 6 4.2.10.2) which included the above lines has been performed. The analysis concluded that systems required to mitigate the consequence of the break will not be impaired by jet impingement or uncontrolled i
w hipping of this line. Restraints, in addition to existing piping and structures, will protect the systems required for safety from being affected by the segment failure.
Certain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the contairunent cooling units, SITS, and their electrical equipment. The containment cooling uni's are designed to maintain their functional integrity following a LOCA (SAR Section 6.2.2.2.2). In addition, all safety injection components located inside the centainment and their associated electrical equipment have been designed to d& stand the LOCA emironment (SAR Section 6.3.2.12.1).11ence, the impact of spatial effects due to a failed segment is assumed to be negligible.
Initia*ing Esent: 1 Initiating Event ID: T6 initiating Event Recoveryt A reactor trip would be initiated due to the RCS pressure drop caused by the segment failure. The required RCS makeup can be provided by the HPSI system once RCS pressure decreases below the shutoff head of the HPSI pumps. Once the letdown isolation valve closes automatically on SlAS to isolate the break, the plant response would be a recovery to normal reactor post trip conditions.
Automatic actuation of the reactor protection system and the appropriate engineered safety features are assumed in response to this initiating event.
Loss of System: S System IPE ID:
Letdown System Recovery: Attempts to recover letdown are not expected to occur because this ponion of the C VCS is not needed for mitigation of this event.
T
FMECA Consequence Information Report Cablemon No. A PEW C4M0H. Rn 00 14 % 91 Pese A19 of A!!
Less of Trals: N Trale ID:
N/A Trale Recovery: !#A Consequence Comment: For the case where the failed segment is isolated, a plant trip will occur due to the initial decrease in RCS pressure. Subsequently, the letdown isolation valve will automatically close upon receipt of SIAS to isolate the failed segment. SIAS is generated by the resulting low pressurizer pressure The plant response will be similar to a typical transient ewnt (i.e., a reactor trip). Per Table I which was developed specifically for ANO 2 using the guidelines in Tables 3.1 and 3.4 of the EPRI l
procedure (EPRI TR 106706), the consequence of the segment failure is MEDIUM.
For the caec where the failed segment remains unisolated, a plant trip will also occur due to decreasing RCS pressure The lost coolant would drain to the containment sump with makeup from the RWT being provided by the IE'SI system once RCS pressure decreases below the shutoff heed of the HPSI pwnps. The consequence. of the unisolated segment downstream of the regenerative heat exchanger inlet valve is bounded by the conesquence upstream of this valve because there is only one valve upstream. Since the consequence downstream of the heat exchanger inlet valve is bounded, only the conesquence of the failed segment between the leidows valve and regeneratin heat exchanger inlet valve is discussed 'The letdown valve must fail in the open position in order for the segment to remain unisolated. According to the IPE, the unreliability of the letdown valve to close is approximately 6.0E.3. The conditional core damage probability of the resulting small LOCA would therefore be reduced by a O
factor of 6.0E.3. Based on Table 1 and the unreliability of the letdown valve, the conditional core damage probability for this isolable small LOCA results in a MEDIUM consequence.
Consequence Category: MEDIUM O'
Co.neg.e.e. Ra.k O
O m
l FMECA - Consequence information Report Cahla'a Na A PENG C4LC 04 R'" 00 14.sep.91 Page A20 of A22 i
Consequence ID: CVCS-C 08 Consequence
Description:
LOCA outside the containment occurs due to a letdown line break during normal power operation.
Break Sin Large Isolability of Hreak: Yes ISO Comments: The break is postulated to occur during normal power operation in the letdown system piping from downstream of containment penetration 2P 14 to upstream of the pressurizer level control valves,2CV-4816 and 2CV-4817. This consequence evaluation includes the welds in the appl' cable portion ofline 2CCD 2 2".
A limidng failure in the segment would cause a rapid decrease in RCS pressure, resulting in a reactor trip and the generation of a Safety injection Actuation Signal (SIAS). Isolation valves inside and outside containment will be closed by SIAS, thus terminating the LOCA. The leakage into the auxiliary building will be stopped at this point in the event. Ilowever, because the break flow would initially be sufnciently large to exceed the normal makeup capacity of 144 gpm, this break would be considered a LOCA. The necessary reactor coolant makeup would be provided by the llPSI system once RCS pressure decreases below IIPSI pump shutoff head following isolation by the SIAS. If necessary, RCS pressure relief um be manually initiated.
Spatial Effects: Local Effected location: Room 2084 Spatial Effects Comments: The subject line segnent is conta!ried wholly within Room 2084 (Upper South Piping Penetration Room) of the Reactor Auxiliary Building. This room has a floor elevation of 360' 0" and the subject pipe segment enters at approximately elevation 362'-0". This room contains numerous safety related components, however, only one of those components is judged to be in close enough proximity (approximately 4') to be in danger of failure due tojet impingement, spray or pipe wlup and this is 2CV4076 2, IIPSI Train "B" cold leg injection to RCS cold leg "2P32D" isolation valve. Because of the SIAS isolation of this line segment following the break, 3
flooding in this room or components at lower evalations is not considered credible.
A loss of this single HPSI injection line valve does not render the "B" IIPSI train inoperable.
Initiating Esent: 1 initiating Esent ID: T6 Initiating Escut Recmery: A reactor trip on low pressure followed in quick succession by a SIAS initi:. tion would occur due to this isolable LOCA. Makeup to the RCS can be provided by the IIPSI system once RCS pressure decreases below the shutoff head of the IIPSI pumps. Following isolation of the LOCA by the SIAS initiation, response of the plant would be a recovery to nonnal reactor post trip conditions.
Automatic actuation of the reactor protection system and the appropriate engineered safety features are assumed in response to this initiating event.
Loss of S stem: S S stem IPE ID:
Letdown 3
3 S stem Recm ery: Recovery of the letdown system would not be a priority because this system is not required for 3
post-trip mitigation.
Less of Train: N Train ID:
N/A Train Recmcry: N/A
FMECA ConsequenceInformation Report C8'alaac..n A reno.cAfoote.h. oo ls-seV91 Pagt All of A!!
Consequence Cosament: For the case w here the f.*iled segment it if 'Nd, a plant trip will occur due to the Liitial decrease in RCS pressure. Subsequendy, tlw letdon isolation valve and containment isolation valves will automatically close upon roccipt of SIAS to isolate the failed segment. SIAS is generated by the resulung low pressurizer pressure. Once the failed segment is isolated, plant response will be similar to a typical transient event (I c., a reactor trip). Per Table I which was developed specifically for ANO 2 using the guidelines in Tables 3.1 and 3.4 of the EPRI procedure (EPRI'Ibl06706), a MEDIUM consequence is assigned for this case.
For the case where the failed segment remains unisolated, a plant trip will also occur due to decreasing RCS pressure. The lost coolant would drain to tlw Reactor Auxiliary Building (RAB) sump with makeup from the RWT being provided by the HPSI system once RCS pressure decreases below the shutoff head of the HPSI pumps. The letdon valve and containment isolation valves must all fail in the open position in order for the segment to remain unisolated. Considering a common fault that would cause all valves to remain open when commanded to close and according to the IPE, the unreliability of these valves to close is approximately 6.0E-4. The conditional core damage probability of this isolable small LOCA would therefore be reduced by a factor of 6.0E 4. Based on Table I and multiple failures that must occur in order for the segment to remain unisolated, the conditional core damage probability for this isolable small LOCA results in a LOW consequence. Considering that the two active barriers to protect against containment bypass must both fail to close in order for the containment to be bypassed, the impact on containment performance results in a O
MEDIUM consequence (Table 3.3 of EPRI procedure TR.106706). Therefore a MEDIUM consequence is assigned for tie cases evaluated.
Consequence Category: MEDIUM O
Co.neque.e na.k O
FMECA Consequorce infonmtion Report Cabubean & AMAGCALC414, /w. 00 lsspin Apr A22 of A22 Talde 1 ASSIGNED CONSEQUPNCE CATEGORIES FOR ANO2INmA11NGEVIMS ineuaues tient inausung teent Lasueuns teent Deeenpuse It CDF CCDP Conseguence ry, (tsyr)
(CDF/tt Preg )
I k<nmare Sung HA N'A N'A NA (verstume Shute.mn
$4 Refuring II Antwipsied kr.ctre Tnp. (T6) 2 03
$ 91E45 2 93E46 k @ nfM (versiianal Cacwience Less of f own Casmesece 8vmarn. U2) 0 25 0 99E47 339E45 MIDIUM Twbane inp. (TI) 0 76 2 27E45 298E M MIDTUM Infrequent less of CW!ane l'own (T3) 5 84E42 172E45 2 9She$
kODRM Leus of SW hav 2P4A. (76) 7 38E42 214E47 2 90E46 MEDIUM
Less of 8WIvnp 2P4B.(19) 13tE42 2ME47 177E46 LGDIUM ai IV Linuting F auha Euessive isedweier. (T4) 9 40E44 187E&
199E46 MIDIUM a Accidenu Sisam LunT esdwater Lane break. Cf 5)
I l0E43 l l3E49 103E45 MIDIUM Totalless of 8W.(T7) 9 45E43 214E4m 392E M 100H Lees dDC hus ?DDI.(TIO) 3 fa4EM 9 80E46 249042 IDOH "
8 Lass of DC bus 2D02 -(Til) 3WE44 I 1110m 281043 100H '
Lens of AC Hun 2A3. (Tit) 3 WE44 323E46 8 20E43 100H*
Lens of AC Bus 2A4. (T13) 3 94E44 3 78E48 I47E M 100H" Less of 480V Land Cenist 2B5. (T14) 104E43 190E47 113E44 IDOH
- Lees of 480VI,nad Cemar 2B6.(Til) 104F43 120047 l l5E44 IDGH "
Small LOCA. (8) 5 00E43 171ba6 3 43E44 IDOH Medium LOCA.(M) 1 OuE43 174E a I 14E43 IDOH 143:LOCA.(A) 100E N 139ti42 139E42 IDOH Swam Osnerstar Tuto kuptwo. (R) 9 77E43 9 54 Eat 9 76544 MEDIUM NOTE:
1.
Thts 6mnasung event resolu in a reacter tity sad less of one trote of serute water to E3F lesda.
2 Tha 6 iu.u s neni mens in a rucier trip.no peru i of.tr.no p.wer.
a na inniauna cent m ns e e runo, trte..d io er p.wer t. e.e trein or nous.une eyite.m d
"lilGll" CCDP > 10
" MEDIUM" 10' < CCDP < 10*
4
" LOW" CCDP < 10 The abose table was developed for the ANO-2 specifle initiating events. It is based on the inferination prwided in Tables 3.1 and 3.4 of the EPRI RISI procedure (EPRI TR.106704 The initiating event descriptions (for event estegories 11.111. A IV) and associated initnettng event frequencies and Core Damage Frequencle (CDFs) were estracted tw Tables 3J4 and 3.14-7A of the ANO-2 IPE (Report 94-R.
200fL01. Res. 0),
O
Calculation No. A PENG CALC 014, Mov. 00 Page 81 of 823 APPENOlX B
'FMECA. OEGRADATION MECHANISMS' (Attachment Pages B1823)
O ABB Combustion Engineering Nuclear Operations
=_ _
""' lad n A*a AMMQN. Rn. 00 FMECA - Degradation Mechanisms l' age B2 of B23 Weld Sastem ID Segment Line Number Line Description Number Weld I. mention T
C P
I M
E F
O CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-001 Weld at RCS cold leg for No No No No No No No No (upstream ofisolation letdown line ulte -inside centainment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-002 Upstream ofelbow #16 No No No No No No No No (upstream ofisolation vahr-inside i
containment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-003 Downstream ofelbow #16 No No No No No No No No (upstream ofisolation uhr -inside contaimnent)
]
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-004 Upstream ofelbow fl7 No No No No No No No No (upstream ofisolation vahr-inside containment)
CVCS CVCS4)01 2CCA-12-2" CVCS letdown line 37-005 Downstream ofelbow #17 No No No No No No No No (upu-n ofisolation uhr -inside containment)
CVCS CVCS4)01 2CCA-12-2" CVCS letdownline 37-006 Upstream ofelbow #18 No No No No No No No No (ups-n ofisolation uht -inside containment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-007 Dv-mu-u ofelbow fl8 No No No No No No No No (upumn ofisolation vaht -inside containment)
Desradstion Mechannms T-Thmnal Fatigue P - Prunary Waser Swess Cesresion Cracking (PWSCC)
M - MicrobeologicmHy inneenced Carramen (MIC)
F-Flow Accelermoed Coms.m C-Cerrosson Cracksng I-LL y Stress Cerronen Cracking (10 SCC)
E-Erossan-C: itsexe 0-Other O
O O
p n
^
G G
)
C"*'"" " h. WWC4/4. Rn. W FMECA - Degradation Mechanisms Page 23 cf B23 W eld System ID Segment Line Number Line Description Neasber Weld Location T
C P
I M
E F
CVCS letdown line 37-008 Upstream ofelbow #19 No No No No No No No No l
(upstream ofisolation uhr-inside containment)
CVCS letdona line 37-009 Danistream orcibow #19 No No No No No No No No (upstream ofisolation uhe -inside containment)
CVCS letdown line 37-010 Upstream ofelbow t20 No No No No No No No No (upstream ofisolation uhr -inside containment)
C /CS CVCS-001 2CCA-12-2*
CVCS letdown line 37-011 Downstream ofcIbow #20 No No No No No No No No (upstream ofisolation vahr -inside contamment)
CVCSletdownline 37-014 Upstrearr of elbow #21 No No No No No No No No (ophm.i ofisolation vahr -inside containment)
CVCS letdown line 37-015 DumoLw.h of elbow #21 No No No No No No No No j
(upstream ofisolation vahr-inside contamment)
CVCS letdown line 37-016 Upstream orcibow #22 No No No No No No No No (upstream ofisolation vahr-inside containment)
Daarneman h j
T-Thermal Fatigue P - Pnreary Waner Stress Carrossan Cradmg (PWSCC)
M-MkrabsolepcmityInfluencedCer essen(MIC)
F Flour Accelerated Cerrones C-Cerrassen Qaciung I-1rsergranular Stress Cerrassen Crudung(IGSCC)
E-Eressen-Cavesnien 0-Oeur r
colarono,r xo. x-Psys.catc_of,, y,v. m ISSep-97 FMECA - Degradation Mechanisms Page B4 of B23 W eld System ID Segment Line Number Line Description Number Weld teention T
C P
I M
E F
0 CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-017 Downstream ofelbow #22 No No No No No No No No (upstream ofisolation l
vahr -inside l
contaimnent)
CVCS letdown line 37 018 Upstream of elbow #23 No No No No No No No No (upstream ofisolation vahr -irside contaimnent)
CVCS CVCS.001 2CCA-12-2" CVCS letdoun tine 37-019 Downstream oreIbow #23 No No No No No No No No (upstream ofisolation vahr -inside containment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-020 Upstream of elbow #24 No No No No No No No No (upstream ofisolation vahr -inside containment)
CVCS CVCS-001 2CCA-I?-2" CVCS letdown line 37-021 Downstream ofcIbow #24 No No No No No No No No (upstream ofisolation vahr -inside containment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-022 Upstream ofelbow #26 No No No No No No No No (upstream ofisolation vahr -inside containment)
CVCS.
CVCS40I 2CCA-12-2" CVCS letdown line 37-023 Downstrer.m of elbow #26 No No No No No No No No (upstream ofisolation uhr-inside containment)
Dearndstem Medunrars T-Timrmal Fatigue P - Pnrnary Water Stress Commenn Cradung (TWSCC)
M-i.fa4 4InnuencedCommsem(MIC)
F-Flow AccelerusedCarrosson C-Commion Cradung I
- e,_a Stress Carramen CrockingOGSCC)
E - Erossan-Cavsanon 0 -Othe O
O O
O O
O C"'"*'n.Va MMC 4M. Rn. 00 FMECA - Degradation Mechanisms Page B5 of B23 Weld System ID Segment Liec Number Line Description Noseber Weld location T
C P
I M
E F
0 CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-024 Upstream of elbow #25 No No No No No No No No l
(upstream ofisolation valw -inside containmen$
CVCS Ictdown line 37-025 Downstream of cibow #25 No No No No No No No No (upuwu ofisolation vaht -inside containment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37426 Upstream of 2" tee #27 No No No No No No No No (up^ wu ofisolation u
valve - inside containment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37427 Downstream of 2* tee #27 No No No No No No No No (upstream ofisolation (drain side) valve-inside containment)
CVCS CVCS-001 2CCA-12-2" CVCS letdown line 37-028 Upstream of manual valve No No No No No No No No (up wiiofisolation 2RC-4A u
valve-inside containment) l CVCS CVCS-001 2CCA-12-2" CVCS Ictdown line 37-031 Downstream 0f 2" tee #27 No No No No No No No No (upstream ofisolation valw -inside containment)
CVCS letdown line 37 032 Upstream of motor-No No No No No No No No m ofisolation operated valve 2CV-4820-2 (up-u i
valve 'nside 1
containment) p== Mechenems T-Thermal Fatigue F - Pnmary Waser Sims Commen Crechng (F%W M - MuroinelegicmEy Imammcel Common (MIC)
F-Fle= Accelermoed Commen c-Comme. Cr. cons
- :.a.,
sire CommariCr congposCC)
E-Erosum-Canimasem O-Odier
l FMECA - Degradation Mechanisms C"#"' "' " #" " " # " 8 " "
l'are B6 of B 3 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
Charging line to RCS 41403 Downstream ofcheck No No No No No No No No cold leg 2P-32C ralve 2CVC-28C l
Charging line to RCS 41-003A Upstream of 2* tee #15 No No No No No No No No cold leg 2P-32C CVCS CVCS-001 2CCA-26-2*
Charging line te RCS 41-C B Downstream of 2* tee #15 No No No No No No No No cold leg 2P-32C CVCS CVCS-001 2CCA-26-2*
Charging line to RCS 41-003C Branch portion of 2* tee No No No No No do No No l
cold leg 2P-32C
- 15 l
Charging linc to RCS 41-003D Upstream ormanual vaht No No No No No No No No cold leg 2P-32C 2CVC-1188 in drain path CVCS CVCS-001 2CCA-26-2*
Chargingline to RCS41-004 Upstream ofelbow #12 No No No No No No No No j
cold leg 2P-32C (ISO 2CCA-26-16)
Charging line to RCS41-005 Dumscam of elbow #12 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-16) l CVCS CVCS-001 2CCA-26-2*
Charging line to RCS41-006 Upstream ofcIbow #11 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-16) l l
Charging line to RCS41-007 Doumstream ofcIbow #11 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-16)
Charging line to RCS41-008 Upstream ofelbow #10 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-16)
Charging line to RCS41-009 Dv-ismu of elbow #10 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-16)
Charging line to RCS41-010 Upstream ofelbow #9 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-16)
Dearadsbon Methernsrns T-Thermal Fatigue F - Pnrnary Water Stress Cermoon Cradang (7%W M - MicrduologicaDy Irinuenced Cerrassco (MIC)
F-flow Accelerused Carmense c-Cerresien Creciung I-Irsergrar=> tar stress Cerrown Cracong 00 SCC)
E - Erassan-Cavitatus 0- ooier O
O O
- O O
O FMECA - Degradation Mechanisms Page B7 of B23 W eld System ID Segment IJee Number Line Description Number Weld 1.mestion T
C F
I M
E F
Charging line to RCS 41411 Domistream ofcibow #9 No No No No No No No No i
cold leg 2P-32C (ISO 2CCA-26-16) i CVCS CVCS-001 2CCA-26-2*
Charging line to RCS41-012 Upstream ofcibow #h No No No No No No No No cold kg 2P-32C (ISO 2CCA-26-16)
Charging linc to RCS 41 013 Domistream oicibow #8 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-16)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS41-015 Upstream ofelbow #14 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
Charging linc to RCS41-016 Du-s-n of elbow #14 No No No No No No No No cold leg 2P-32C OSO 2CCA-26-15)
Charging line to RCS41-017 Upstream ofcibow #13 No No No No Pb No Na No cold leg 2P-32C OSO 2CCA-26-15)
Charging line to RCS41-018 Domistream ofelbow fl3 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
Charging line to RCS41-019 Upstream ofcibow #12 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
Charging line to RCS41-020 Dv->&
is ofelbow #12 No No Ne No No tb No No cold leg 2P-32C (ISO 2CCA-26-15)
Charging line to RCS41-021 Upstream ofcibow #11 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
Chargingline to RCS41-022 Domistream ofcibow #11 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
Charging line to RCS41-023 Upstream ofelbow #10 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
Desadsnian Mechan==s T rimensiratigue r-Pnmarywesersirescerro ancrachns(rwscc)
M-N rammencedcarre (unc)
F-mm. Acceserusedcorre=en c-cerro creciung I-:
Sarns Commean Cradmg(IGsCC)
E - Erosion-Ca#ssean 0-Oeser
FMECA - Degradation Mechanisms Calculanon A MMWI4 Rn 00 Pop BS of B23 Weld System ID Segment Line Number Line Description Number Weld Lacetien T
C P
I M
E F
0 CVCS CVCS-001 2CCA-26-2" Charging linc to RCS41-024 Dowr. stream of elbow #10 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS 4I-025 Upstream ofelbow #9 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS41-026 Downstream ofelbow #9 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS41-027 Upstream ofelbow #8 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
CVCS CVCS-001 2CCA-26-2" Charging linc to RCS41-028 Downstream ofelbow f8 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-15)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS41-030 Upstream ofcibow #9 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS41-031 Dumswiiofcibow #9 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS41-032 Upstream ofelbow #3 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
CVCS CVCS-001 2CCA-26-2" Charging line to RCS41-033 Downtream ofelbow #3 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
Charging line to RCS 4I-034 Upstream ofcibow #7 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
Charging linc to RCS41-035 DowTistream orelbow #7 No No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
CVCS CVCS-001 2CCA-27-2" Charging line to RCS40-005 Dums-n ofcheck No No No No No No No No cold leg 2P-32B valve 2CVC-28B Desradaten Medutrasms T-nermal Fatigue F - Frunary Water Stress Common Creding (P%W M-Mxrebiologicany haflamced carre sen(MIC)
F-How Acceleruerd Common c-Commen Crochng I-:.L.,_._lar stres Commen Cracking posCC)
E - Eremen -Cavstshee 0-Other O
O O
3 3
pJ (G
(J
- "d*" #" N N -0 4.R m 00 FMECA - Degradation Mechanisms l' axe B9 of B23 W eld Systems ID Segment Une Number une Description Nemter Weld lacaties T
C F
I M
E F
Chargingline to RCS 40406 Upstream of2" tee #12 No No No No No No No No cold leg 2P-32B OSO 2CCA-27-3)
Charging line to RCS40-007 Downstream of 2* tec #12 No No No No No No No No cold leg 2P-32B (ISO 2CCA-27-3)
Charging linc to RCS40-008 Upstream orelbow s7 No No No No No No No No cold leg 2P-32B (ISO 2CCA-27-3)
Charging line to RCS40-009 DowTtstream ofcibow #7 No No No No No No No No cold leg 2P-32B 050 2CCA-27-3)
Charging line to RCS40-010 Douvistream ofpiping No No No No No No No No cold leg 2P-32B segmer* #1 OSO 2CCA-27-3)
Charging linc to RCS40-011 Upstream ofpiping No No No No No No No No cold leg 2P-32B segment #4 OSO 2CCA-27-2)
Charging linc to RCS40-012 Upstream ofelbow f7 No No No No No No No No cold leg 2P-32B OSO 2CCA-27-2)
Charging line to RCS40-013 Dumoumik of elbow #7 No No No No No No No No cold legI?-32B (ISO 2CCA-27-2)
Charging line to RCS40-014 Upstream ofelbow #6 No No No No No No No No cold leg 2P-32B (ISO 2CCA-27-2)
Chargingline to RCS40-015 DowTistream orcibow #6 No No No No No No No No cold leg 2P-32B (150 2CCA-27-2)
Chargingline to RCS40-016 Upstream ofcibow #5 No No No No No No No No cold leg 2P-32B OSO 2CCA-27-2)
M m M %====
T.Thenal Fatisme P-Pnmary Water Stras Cammsen Crming (PRM M - Mandadopcally bdlemced Cerressan (MIC)
F-flour Accelerased Cerrassam c-Commsen Craime I-beeryanaser sans Commsan Cracting OOSCC)
E-Eremen-Cavemanna 0-osier
~
1.
'*7 Caladan n Na MMC 4U. Ra. 00 FMECA - Degradation Mechanisms E' age B10 of B23 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
O CVCS CVCS401 2CCA-27-2" Charging linc to RCS40-017 Dennstream ofelbow #5 No No No No No No No No cold leg 2P-32B (ISO 2CCA-27-2)
CVCS CVCS-001 2CCA-27-2" Charging line to RCS 40-0iR Upstream of elbow #7 No No No No No No No
?4 cold Icg 2P-32B (ISO 2CCA-27-1)
Charging line to RCS40-019 Dounstream oreR wr #7 No No No No No No No Ne cold leg 2P-32B (150 2CCA-27-1)
Charging linc to RCS40-020 Upstream ofelbow #6 No No No
?k No No No No cold leg 2P-32B (ISO 2CCA-27-1)
Charging line to RCS40-021 Downstream ofelbow #6 No No No Ne No No No No cold leg 2P-32B (ISO 2CCA-27-1)
Charging linc to RCS 40433 Downstream of 2* tee #19 No No No No No No No No cold leg 2P-32B (on <' rain side, ISO 2CCA-27-3)
CVCS CVCS-001 2CCA-27-2" Charging linc to RCS40-034 Downstream oreIbow f22 No No No No No No No No cold leg 2P-32B (ISO 2CCA-27-3)
CVCS CVCS-001 2CCA-27-2" Charging line to RCS 40435 Upstream of manaul valve No No No No No No No No cold leg 2P-32B 2CVC-1186 in drain path CVCS CVCS-002 2CCA-26-2*
Charging linc to RCS41-036 Upstream orelbow #6 Yes No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
Charging line to RCS41-037 Downstream ofcibow #6 Yes No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
CVCS CVCS-002 2CCA-26-2" Charging linc to RCS41-038 Upstream ofelbow #5 Yes No No No No No No No cold leg 2P-32C (ISO 2CCA-26-14)
Dearadation Mecharnoms T-Thermal Fatigue P - Pnmary Water stress Common Cradmg (PWsCC)
M - MxrobeowgicaDy influenced Commen (MIC)
F-Flow AccelerseedComme c-C rro onCr ctmg I-traergr== tar sve Com.m Cr.ct.: poscr,
E-Eroman - Cavstation 0-Otler O
O O
%m q
J Calculatum No A-PENG-CALC-014. Rev. 00 i4-sym FMECA - Degradat. ion Mechannsnis e,ge sii of s2s
^
W eld System ID Segment Line Number Line Description Number Weld Leestion T
C F
I M
E F
Charging line to RCS41-039 Dv.obum ofcIbow #5 Yes No No No No No No No cold leg 2P-32C at RCS cold leg 2P32C (ISO 2CCA-26-14)
Chargingline to RCS41-040 At RCS coldleg 2P32C Yes No No No No No No No cold leg 2P-32C nonle CVCS CVCS402 2CCA-27-2*
Charging line to RCS40-022 Upstream ofcIbow #5 Yes No No No No No No No cold leg 2P-32B (ISO 2CCA-27-l)
Charging line to RCS40-023 Downstream of cibow #5 Yes No No No No No No No cold leg 2P-328 (ISO 2CCA-27-1)
Chargingline to RCS 40424 Upstream ofelbow #4 Yes No No No No No No No cold leg 2P-32B (ISO 2CCA-27-1)
Charging line to RCS40-025 Dommstream ofcibow 84 Yes No No No No No No No cold leg 2P-32B at RCS cold leg 2P32B (ISO 2CCA-27-1)
Chargingline to RCS40-026 Weld at RCS coldleg for Yes No No No No No No No cold leg 2P-32B charging line CVCS CVCS-003 2CCA-26-2*
Charging linc to RCS41-001 Dom 1: stream of motor No No No No No No No No cold leg 2P-32C operated vaht 2CV-4831-2 CVCS CVCS-003 2CCA-26-2*
Charging line to RCS41-002 Upstream ofcheck raht No No No No No No No No cold leg 2P-32C 2CVC-28C CVCS CVCS-003 2C'A-27-2*
Charginglinc to RCS40-001 Dv-iob-. of motor-No No No No No No No No cold leg 2P-32B operated vahr 2CV-4827-2 CVCS CVCS-003 2CCA-27-2*
Chargingline to RCS40-002 Upstream of 2* tec #9 No No No No No No No No cold leg 2P-32B (ISO 2CCA-27-3)
Desredema Mechanens T-Thennel reigne r-Prunary weier sirees cerroe== ca.ckW (rrscc) u - Micretacie,cass, kolmenced corree.e. (usC) r-rio. Acc ier dcorre c-carreann onckins I - :.e,,
- strees correens cracksng(10 SCC)
E-Eros.on -Cavannen 0-Odier
FMECA - Degradation Mechanisms N'"'"*"' A'oMMM'414, Rm 00 Page Bl2 of B23 W eld Sjstem ID Segment Line Number Line Description Number Weld Iscation T
C P
1 M
E F
Charging line to RCS40-003 Upstream ofcheck nht No No No No No No No No cold leg 2P-32B 2CVC-28B CVCS CVCS-003 2CCA-27-2" Charging line to RCS40-027 Upstream of 2" tec #9 No No No No No No No No cold leg 2P-328 (shown on ISO 2CCA 3)
Charging line to RCS40-028 Downstream ofc! bow #4 No No No No No No No No cold leg 2P-32B (ISO 2CCA-27-4)
CVCS CVCS-003 2CCA-27-2" Ch.rging line to RCS40-029 Downstream of manual No No No No No No Nc No cold leg 2P-32B nlw 2CVC-27 CVCS CVCS-003 2CCA-27-2" Charging linc to RCS40-031 Upstream ormanual nive No No No No No No No No cold leg 2P-32B 2CVC-27 CVCS CVCS-003 2CCA-27-2" Charging line to RCS40-032 Downstream of 2* x 1*
No No No No No No No No cold leg 2P-32B reducer #8 (2* side -ISO 2CCA-27-4)
CVCS letdown line 37-029 Dumohcam of manual No No No No No No No No (upstream ofisolation uht 2RC-4A vaht -inside containment)
CVCS letdown line 37-030 Upstream of manual vaht No No No No No No No No (upstream ofisolation 2RC-5A ulve -inside containment)
CVCS Ictdown line 37-033 Downstream of moter-No No No No No No No No (upstream ofisolation operated vahr 2CV-4820-2 valve -inside containment)
Deuradstxm Mechenums T-N Fatigue P - Prunary Waner stress Corressen Crecisig trum M - Meret=alogicany Innerne:d Commen (MIC)
F-Flow AmeierenedCommeu C-Common Chisig I a,_A Sems Commen Crachng(1GSCC)
E-Eremen-Cavnence 0 -Odur O
O O
'N' FMECA - Degradation Mechanisms C"'*" ##"""'-8" "" 88 l' age B13 of B23 W eld System ID Segment Line Number Line Descripties Number Weld IA= wies T
C P
I M
E F
0 CVCS CVCS-004 2CCA-12-2' CVCS letdown line 37433A Downstream of 2* x 2* x No No No No No No No No fopstream ofisolation 3/4* reducing tec #32 nht -inside containment)
CVCS letdosm line 37.n34 Upstream ofmotor-No No No No No No No No (upstream ofisolation operated waht 2CV-4821-1 nht -inside containment)
Charging line to RCS 41-003E 12v=.h.. of manual No No No No No No No No cold leg 2P-32C vaht 2CVC-1188 in drain Path CVCS-CVCS-004 2CCA-26-2*
Charging line to RCS 41-003F Upstreamofmanualnhe No No No No No No No Ne cold leg 2P-32C 2CVC-1189 in drain path CVCS CVCS-004 2CCA-26-2*
Chargingline to RCS 41-003G Dominstream of manual No No No No No No No No cold leg 2P-32C vahr 2CVC-1189 in drain Path CVCS CVCS-004 2CCA-27-2*
Charging line to RCS40-036 Dom 1:stremm of manual No No No No No No No No cold leg 2P-32B vaht 2CVC-1186 in drain Path CVCS CVCS-004 2CCA-27-2*
Charging line to RCS40-037 Upstream ofmanual vaht No No No No No No No No cold leg 2P-32B 2CVC-1187 in drain path CVCS CVCS-004 2CCA-27-2*
Charging line to RCS40-038 Dominstream of manual No No No No No No No No cold leg 2P-328 nhr 2CVC-1187 in drain Pd Desradshan Mechensen T-Thermal Fangue P-Pensry Water Stress Common Craiing (PWSCC)
M-MZ
.S i nnuencedCommmen(MIC)
F-Flow AccelerasedCarrwesen I
C-Carromaan Chiing I - Innergreenlar Stres Commsen Craciung (IGSCC)
E-Eronen-Cansmian 0-Odier
i w 97 FMECA - Degradation Mechanisms Calculatiam No. A-PEVG CALC-014. Rev. 00 Page B14 of B23 W eld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
0 CVCS CVCS-004 2CCB-1-2" Letdown line - 2*
37-063 Downstream of motor No No No No No No No No portion trpstream of operated wahr 2CV-432I-regenerative heat I (ISO 2CCB-1-1) exchanger CVCS CVCS-004 2CCB-I-2" Letdown line - 2"37-064 Upstream of 4*r.2*
No No No No No No No No portion upstream of eccentric reducer #12 (ISO regeneratht heat 2CCB-1-1) exchanger CVCS CVCS-004 2CCB-1-21/2" Letdown line - 21/2"37-073 Downstream of 4*x21/2*
No No No No No No No No portion upstream of eccentric reducer #4 (ISO regenerative heat 2CCB-1-1) exchanger CVCS CVCS-004 2CCB-I-21/2" Letdown line - 21/2"37-074 Upstream orelbow f2 No No No No No re, No No portion upstream of (ISO 2CCB-1-l) regenerathe heat exchanger CVCS CVCS-004 2CCB-1-21/2" Letdown line - 21/2"37-075 Downstream of elbow #2 No No No No No No No No portion upstream of (ISO 2CCB-t-l) regenerathe heat exchanger CVCS CVCS-004 2CCB-1-21/2" Letdown line - 21/2"37-076 Downstream of piping No No No No No No No No portion upstream of section #1 (ISO 2CCB-1-1) regenerative heat exchanger CVCS CVCS404 2CCB-1-4" Letdown line - 4*
37-065 Dumahmt of 4"x2" No No No No No No No No portion upT. of eccentric r-ducer #12 (ISO regenerathe heat 2CCB-I-1) exchanger Dearedsami Mecharmms T IhermalFatigue P-Pnmary Wster seress Corrassen Cracking (PW3CC)
M - MurcimologicmDy Iremenced Cermusen (MIC)
F-Flow AccelerseedCommsen C-Common Crackmig I-Inse granuter Stress Common Cracking OGSCC)
E-Eresums -Cavitatsen 0 -Other O
O O
O O
O C"' #""' A'" N"C4N R" "
FMECA - Degradation 5'.-chanisms Page B15 of B23 W eld Systen ID Segment Line Number Line Description Number Weld Locaties T
C P
I M
E F
Letdown line - 4*
37-072 Upstream of 4*x21/2*
No No No No No No No No portion upstream of eccentric reducer #4 (ISO regenerative heat 2CCB-1-1) exchanger CVCS CVCS-004 2CCB-14*
letdown line - 8*
37-066 Dv-im> =..of8"x4*
No No No No No No No No portion upsc=T. of eccentnc rer'socr #11 (ISO regenerative heat 2CCB-1 ')
exchanger CVCS CVCS-004 2CCB-14" Letdown line - 8"37-067 Upstrezm orcibow f9 No No No No No No No No portion upstream of (ISO 2XB-t-1)
.c w dive heat s
exchanger CVCS CVCS-004 2CCB-14*
Letdown line - 8'37-068 Pamistream of elbow #9 No No No No No No No No portion upstream of (ISO 2CCB-1-1) regenerathe heat exchanger CVCS CVCS-004 2CCB-14" Letdown line - 8'37-069 Upstream ofcibow #7 No No No No No No No No portion ups-u of (ISO 2CCB-1-1) regenerathe heat exchanger CVCS CVCS-004 2CCB-14" Letdown line - 8"37-070 Dv-me-- of elbow #7 No No No No No No No No portion uym-- of (ISO 2CCB-I-1) regenerathe heat exchanger CVCS CVCS-004 2CCB-14" Ixtdown line - 8"37-071 Upstream of 8"x4*
No No No No No No No No portion uyswT. of eccentnc reducer #5 (ISO regenerathe heat 2CCB-1-1) exchanger Desredstma Medunens T-Thenal Fatigue P - Frunary Waser s:res Cerreman Cracksng (F%W M - Macrebesleycmay Isamenced Canomes (M3C)
F-Flow AccelermoedCarresnese C-Careman Cracking I-beeryanular seren Commson Crackirig (IGsCC)
E-Eressen -Cavemenw O-Oswr
FMECA - Degradation Mechanisms C"#"" " Na ANWQu. Rw. 00 Page B16 of B23 W eld System ID Segment Line Number Line Description Number Weld Imestica T
C P
I M
E F
letdown line - 2*
37-078 Downstream of 21/2*x2" No No No No No No No No portion downstream of concentric reducer #42 regenerathe heat (ISO 2CCB-2-1) exchanger CVCS CVCS-004 2CCB-2-2*
lxtdown line -2"37-079 Upstream of 2* tec #34 No
?4 No No No te No No portion downstream of (ISO 2CCB-2-1) regenerathe heat exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2."
374)80 Downstream cf 2* tee #34 No No No No No No No No portion downstream of (ISO 2CCB-2-1) regenerathe heat exchanger CVCS CVCS-004 2CCB-2-2*
letdown lim:- 2*
37-081 Upurcam of 2* tee #32 No No No No No No No No portion downstream of (ISO 2CCB-2-1) regenerathe heat exchanger CVCS CVCS-004 2CCB-2-2*
Iendown line - 2*
37-082 Upstream of 2*xl*
No No No No No No No No portion downstream of reducing insert #33 (ISO Kgmuu the heat 2CCB-2-1) exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2*
37-083 Downstream of 2* tee #32 No No No No No No No No portion dumsiezuii of (ISO 2CCB-2-1) regenerathe heat exchanger CVCS-CVCS-004 2CCB-2-2*
Letdown line - 2*
37-084 Upstream of 2*xt*
No No No No No No No No portion downstxam of reducing insert #35 (ISO regenerathe heat 2CCB-2-l) exchanger Desradatma Methernsms T-ThermalFati ue P-Prwnery Water Stress Cormuon Cracing (PWSCC) t M - M'cruteologienDy Irdleenced Cemsson (MIC)
F-Tkm AccelerasedCommon a
C-Commen Crackms
!-Insergr== tar stress Common Oracing OGSCC)
E-Eramos -Cavastma O-Oeur O
O O
D O
O C"'"'""' 'Va MMK-ON. Rm 00 FMECA - Degradation Mechanisms Page BIT of B23 W eld Systens ID Segment Line Number Line Description Number Weld leestion T
C F
I M
E F
Letdown line - 2*
374185 Upstream orelbow s31 No No No No No No No No portion downstream of (ISO 2CCB-2-1) icgcaaths heat exchanger CVCS CVCS-004 2CCB-2-2*
Ixtdown line -2*
37-086 Downstream ofcibow #31 No No No No No No No No por: ion downstream of (ISO 2CCB-2-1) icguaatht heat exchanger CVCS CVCS-004 2CCB-2-2*
letdown line-2*
37-087 Upstream ofelbow s30 No No No No No No No No porten downstream of (ISO 2CCB-2-1)
. 3saathe heat exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2"37-088 Downstream orelbow f30 No No No No No No No No porten downstream of (ISO 2CCB-2-1) l
.sgam.the heat exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2*
374)89 Upstream ofelbow #29 No No No No No No No No porten dui..aism of (ISO 2CCB-2-1) regeneraths heat exchanger CVCS CVCS-004 2CCB-2-2*
lxtdown line -2*
37-090 D
- ofelbow #29 No No No No No No No No porten downstream of (ISO 2CCB-2-1) icgsmi.t heat exchanger D
CVCS-004 2CCB-2-2" Letdown line - 2*
37-091 Upstream of 2* coupling No No No No No
,No No No porten downstream of
.s
~ 4 s heat exchanger Desredman N "di amencedCarresen(MIC)
F-Fles AcelerusedCarreuen k
T-ThermalFenigue F - Pnmary Waner stress Corressen Crnching (FWSCC)
M'C c-Corremen cruiing 1-a.,
- sims Cerve=en Chdung(IGSCC)
E-Eremen-Caveensam O-OWier
C"'#*'"" " #* A * " C# " # " 88 FMECA - Degradation Mechanisms Page BIS of B23 W eld System ID Segment Line Number Line Description Number Wdd Ementica T
C P
I M
E F
letdown line - 2*
37-092 Dowa<tream of 2*
No No No No No No No No portion downstream of coupling #28 (ISO 2CCB-regenerati r heat 2-1) exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2*
374)93 Upstream of tee #25 (ISO No No No No No No No No portion downstream of 2CCB-2-1) regenersthe heat exchanger CVCS CVCS-004 2CCB-2-2*
letdown sine - 2*
37-094 Upstream of 2*x3/4*
No No No No No No No No portion downstream of reducing insert #26 (ISO regeneratht heat 2CCB-2-1) exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2*
37-095 Downstream of te= #25 No No No No No No W
No portion downstream of (ISO 2CCB-2-1) regenerathe heat exchanger CVCS CVCS-004 2CCD-2-2*
Letdown line - 2*
37-096 Upstream of 2* coupling No No No No No No No No portion downstream of
Letdown line - 2*
37-097 Downstream of 2*
No No No No No No No No portion downstream of coupling #24 (ISO 2CCB-regenerathe heat 2-1) exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2*
37-098 Upstream of c! bow #4 No No No No No No No No portion downstream of (ISO 2CCB-2-1) regenerathe heat exchanger Desradeem Medhannsm:
T-Thermal Fatigue P - Pnmary Water Stress Comman Cracksng (FWSCC)
M - Mscreteolopcally Innuenced Common (MIC)
F.How Aaelerused Cervsnen C-Cer esionCrecing I-beerpinular sire = Common Crock =g OGSCC)
E-Eresum-Cavastman 0-Ot!wr O
O O
~
s s
i FMECA - Degradation Mechanism s C**'"""' A'*"MWM R" 88 Page B19 of B23 1
W eld j
System ID Segment Line Number Line Descriptica Number WeW IAcation T
C P
I M
E F
0 l
Letdown line - 2*
37-099 Dv-nstream orelbow W No No No No No No No No l
portion downstream of (ISO 2CCB-2-1) regeneratist heat cx ger 1
Letdown line - 2*
37-100 Upstream ofelbow #3 No No No No No No No No portion downstream of (ISO 2CCB-2-1)
K s m.m.in t heat exchanger j
Letdown line - 2*
37-101 Downstream ofelbow f3 No No No No No No No No portion e-macam of (ISO 2CCB-2-1) regenerathe heat exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2*
37-102 Upstream ofelbow #2 No No No No No No No No portion du..obme of (ISO 2CCB-2-1) regenerathe heat i
exchanger CVCS CVCS-004 2CCB-2-2*
Letdown line - 2*
37-103 Dv-. sum ofelbow #2 No No No No No No No No portion downstream of (ISO 2CCB-2-1) regeneraths heat exclunger CVCS CVCS-004 2CCB-2-2*
letdown line - 2*
37-104 Dv-.s n of piping No No No No No No No No portion downstream of section #5 at contanament Kgci.u.iht heat r==.ii 2P-14 (ISO
]
exchanger 2CCB-2-1)
CVCS.
CVCS-004 2CCB-2-2*
leidown line - 2"37-105 Downstream of 2*
No No No rio No No No No portion du->su.ni of coupling #34 at K gm ~.ihr heat cotainment peian 2P-exchanger 14 (ISO 2CCB-2-2)
Dewedse am Med====
T-Thermal Fatigue P-Pnmary Water Stress Corrensen Crecitmg(PHW M-L 2
!_ My Induenced Commeen(noC)
F-Fleur Accelermed Carnanni C-Carro cv.ckmg
-senergrannt rswseCerro Cr at-g(IGSCC)
E-Eressan-Cenemasse 0-Odier
Caladatow No_ AMMCW. Rm 00 FMECA - Degradation Mechanisms Page B:0 of B23 Weld System ID Frgment Line Number Line Description Number WeldIsation T
C P
I M
E F
0 CVCS CS-004 2CCB-2 12tdown line - 2*
37-106 Upstream erelbow #1 No No No No No No No No portion downstream of (ISO 2CCB-2-2) regenerathe heat exchanger CVCS CVCS4X)4 2CCB-2-2*
Letdown line - 2*
37-107 Dounstream orelbow #1 No No No No No No No No portion downstream of (ISO 2CCB-2-2) regeneratist heat exchanger CVCS CVCS-004 2CCB-2-2*
1xtdown line - 2*
37-108 Upstream of elbow f2 No No No No No No No No port;on dmmstream of (I O 2CCB-2-2) regeneraQt heat exchanger CVCS CVCS4X)4 2CCB-2-2*
Letdown line - 2*
37-109 Downstream ofcibow #2 No No No No No No No No poruon downstream of (ISO 2CCB-2-2) regenerative heat exchanger CVCS CVCS4)04 2CCB-2-2*
Letdown line - 2*
37-110 Upstream of 2* tee #31 No No No No No No No No portion downstream of (ISO 2CCB-2-2) regeneratist heat exchanger CVCS CVCS M 2CCB-2-2*
Letdown line - 2*
37-111 Upstream of 2*xl*
No No No No N
No No No portion downstream of reducing insert #29 (ISO regeneratisc heat 2CCB-2-2) exchanger CVCS CVCS4)04 2CCB-2-2*
Lerdown line - 2*
37-113 Dewusu-of 2* tee #31 No No No No No No No No portion downstream of (ISO 2CCB-2-2) regenerathe heat exchanger Deeradstum Mechannms T-Thermal Fatigue F - Pnenary Weier Stress Corremen Creeng (fum M-M~J !Afhalsenced Camemmi(30C)
F-fle= AccekrusedCommes c-Common Cracting I-beergranular Stress Correare Cre&ng OGSCC)
E - Enema-Cneatum O-Odier O
O O
y p
(
G
~
C"'c"'" "'" ^'* N " C * ^ " "
FMECA - Degradation Mechanisnes l' age M1 of B23 WeW System ID Segment Line Number Line Descripties Nesser WeW teesties T
C P
I M
E F
Ixtdown line - 2*
37-I14 Upstream of 2* tee #32 No No No No iJo No No No portion downstream of (ISO 2CCB-2-2) regeneratne heat exchanger CVCS CVCS-004 2CCB-2-2" Letdemn line - 2*
7-113 Dumsms. of 2* tec #32 No No No No No No No No portion downstream of OSO 2CCB-2-2)
.s g. ~.G.c heat exchange-CVCS CVCS-004 2CCB-2-2" letdoui line -2*
37-116 Dominstream of 2* tee #32 No No No No No No No No porten Ju naream of (ISO 2CCB-2-2)
.sgsm.sG t heat exchanger CVCS CVCS M 2CCB-2-2" lxtdown line -2*
37-117 Upstream orair operated No No No No N,
No No No portion Jumsw-. of vahr 2CV-4523-2 USO ice.~siive heat 2CCB-2-2) exchanger CVCS CVCS-004 2CCB-2-2" letdown line -2*
37-Il7A Downstreamofair No No No No No No No No portion downstream of operated vahr 2CV-4823-
.sgs~.G t heat 2 (ISO 2CCB-2-2) exchanger CVCS CVCS 004 2CCB-2-2*
Ixtdown line - 2*
37-118 Upstream of 2* tec #33 No No No No No No No No portson downstream of (ISO 2CCB-2-2)
.cs s ~..G.e heat exchanger CVCS CVCS-(44 2CCB-2-2" Letdown line - 2"37-119 Dom 1tstream of 2* tee #33 No No No No No No No No porten downstream of (ISO 2CCB-2-2) icgs~.iive heat CxChaDger Desude== Mect==
T-Therwist Fa6gue P - Pnenary Ws:er Sires Ceneman Cradung(PWSCC)
M-ML-
_ _, hafhsenced Camm.en (MIC)
F-flew AccelermedCarrommese c-Cem c Creasis t - beerarinuter sires Ceminen Crudag (1GSCC)
E-Eramen-Cm O-Other w
e C"' "" " #^ #S"G" #" 88 FMECA - Degradation Mechanisms Page B22 of B23 Weld System ID Segment Line Number Line Description Number Weld Instion T
C P
I M
E F
0 CVCS CVCS-004 2CCB-24" Letdown line - 2"37-120 Upstream ormanual vaht No No No No No No No No portion downstream of 2CVC-I A (ISO 2CCB-2-2) regenerative heat exchanger CVCS CVCS-004 2CCD-2-2" Letdown line - 2"37-121 Downstream of 2* tee #33 No No No No No No No No portion downstream of (ISO 2CCB-2-2) i l
regenerative heat exchanger CVCS CVCS-004 2CCB-2-2" Letdown line - 2"37-122 Upstream orelbow #3 No No No No No No No No l
portion downstream of (ISO 2CCB-2-2) regenerative leat exchanger CVCS CVCS-004 2CCB-2-2" Letuonn line - 2"37-123 Downstream or cibow #3 No No No No te No No No portion downstream of (ISO 2CCB-2-2) regenerative heat exchanger CVCS CVCS-094 2CCB-2-2" Letdown line - 2"37-124 Upstream ormanual vaht No No No No No No No No portion downstream of 2CVC-1B (ISO 2CCD-2-2) regeneratist heat exchanger CVCS CVCS-004 2CCB-2-2" Letdown line - 2"37-125 Downstream of manual No No No No No No No No portion downstream of valve 2CVC-1B (ISO regenerative heat 2CCB-2-2) exchanger CVCS CVCS-004 2CCD-2-2" Letdown line - 2"37-126 Upstream ofelbow #4 No No No No No No No No portion downstream of (ISO 2CCB-2-2) regeneratist heat exchanger Deeradation Mectwrisms T-Thermr.1 Fatigue P - Pnmary Water St.ess Corrosion Cracking (PWSCC)
M - Microteologically Innuenced Corresen (MIC)
F-Flow AcceleratedCarrosion C - Carrosion Cracking I-Intergranular Stress Carrosion Cracking (IGSCC)
E - Eronen-Cavitation 0 - Other O
O O
y V
N'"'""*" #" A *""" R'" "
"~ #
FMECA - Degradation Mechanisms Page B23 of B23 Weld System ID Segment Line Number Line Description Number Weld Location T
C P
I M
E F
0 CVCS CVCS-004 2CCB-2-2" Letdown line - 2"37-127 Downstream ofelbow #4 No No No No No No No No
^
ponion downstream of (ISO 2CCB-2-2) regenerative heat exchanter CVCS CVCS-004 2CCB-2-2" Letdown line - 2"37-128 Upstream of air operated No No No No No No No No portion downstream of valve 2CV-4816 (ISO regeneratirc heat 2CCB-2-2) excLanger 2CCB-2-2" Letdown line - 2"37-129 Downstream of manual No No No No No No No No CVCS CVCS-004 portion downstream of valve 2CVC-1 A (ISO regenerative heat 2CCB-2-2) exchanger CVCS CVCS-004 2CCD-2-2" Letdown line - 2"37-130 Upstream ofelbow #7 No No No No No No No No portion downstream of (ISO 2CCB-2-2) regeneratise heat exchanger CVCS CVCS-004 2CCB-2-2" Letdown line - 2"37-131 Downstream o'clbow #7 No No No No No No No No portion downstream of (ISO 2CCB-2-2) regeneratise heat exchanger CVCS CVCS-004 2CCB-2-2" Letdown line - 2"37-132 Upstream of air operated No No No No No No No No portion downstream of vahe 2CV-4817 (ISO regenerative heat 2CCB-2-2) exchanger CVCS CVCS-004 2CCB-2-21/2" Letdown line - 21/2"37-077 Upstream of 21/2"x2" No No No No No No No No portion downstream cf concentric reducer #42 regenerative heat (ISO 2CCB-2-1) exchanger Deeradation Mechanums T-Thermal Fatigue P - Pnmary Water Stress Cerrosen Cracking (PWSCC)
M-ML"
':M, Innuenced Cerrassen (MIC)
F-flow AccelerusedCceramen Ca cerrosion Cracking 1 - Irmersranular Stress Carrossen Crackmg 00 SCC)
E - Eressen -Cavitahan 0 - Other 1
l
Calculation No. A PENG CALC 014, Rev, 00 Page C1 of C5 O
APPENDIX C
'FMECA - SEGMENT RISK RANKING REPORT *
(Attachment Pages Cl C5)
O ABB Combustion Engineering Nuclear Operations
l} S C='c='aao
- A-PEmcur-*. F' oo
'N" FMECA - Segment Risk Ranking Report raza a era Degradation l
Number Lines in Welds in Degradation Degradation Mechanism Consequence Risk Risk Segment ID of Welds Segment Segment Mechanisms Group ID Category Category Category Rank l
CVCS-001 83 2CCA-12-2",37-001,37-002,37-CVCS-N NONE IIIGH CAT 4 hEDIUM l
003,37-004,37-005, 2CCA-27-2"37-006,37-007,37-008,37-009,37-010, l
37-0II,37-014,37-l 015,37-016,37-017, l
37-018,37-019,37-020,37-021,37 022, 37423,37-024,37-025,37-026,37 4 27,37-028,37-031,37-032//41 003,41-003 A,41-003 B. 41-003C 4I-003D,41-004,41 005,4I406,41-007,41 008,41-009,41-010,41-01I,41-012,41-013,41-015,41-016,41-017,41-018,41-019,41-020,41 021,41-022,41-023,41-024,41-025,41-026,41-027,41-028,41-030,41-031,41-032,41-033,41-034,41-035//40-005,40406,40 407,40-008,40-009,40-010,40-011,40-012,40-013,40 014,40-015,40-016,40-017,40-018,40-0!9,40-O O
O
O O
O Ch &m hAE GCAC&.RnW FMECA - Segment Risk Ranking Report res a 4a l
Degradation Number Lines in Welds in Degradation Degradation Mechanism Consequence Risk Risk Segment ID ofWelds Segment Segment Mechanisms Group ID Category Category Category Rank 020,40-021,40 433,40-034,40-035 CVCS-002 10 2CCA-26-2",41-036,41-037,41-T CVCS-T SMALL IBGH CAT 2 HIGH 2CCA-27-2" 038,41-039,41-040 LEAK
//40-022,40-023,40-024,40-025,40-026 CVCS-003 10 2CCA-26-2",41-001,41-002//40-CVCS-N NONE LOW CAT 7 LOW 2CCA-27-2" 001,40-002,40-003,40-027,40-028,40-029,40-031,40 4 32
1 14-Sep-97 FMECA - Segment Risk Ranking Report C"k"'"""""* d'7j'c';","c",
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 CVCS-004 81 2CCA-12-2",37-029,37-030,37-CVCS-N NONE MEDIUM CAT 6 LOW 2CCA-26-2",
033,37-033 A, 37-2CCA-27-2",
034 // 41-003E,4I-2CCB-1-2",2CCD- 003F,41-003G // 40-l-21/2",2CCB 036,40-037,40-038 4", 2 CCB-1-8",
//37-063,37-064 //
2CCB-2-2",2CCD-37-073,37-074,37-2-2 I/2" 075,37-076//37-065,37-072 //37-066,37-067,37-068, l
37-069,37-070,37-071//37-078,37-079,37-080,37 081,37-082,37-083,37-084,37-085,37-086,37-087.37-088,37-089,37-090,37-091,37-092,37-093,37-094,37-095,37-0 %,37-097,37-098,37-099,37-100,37-101,37-102,37-103,37-104,37-105,37-106, 37-107,37-108,37-109,37-110,37-111,37-113, 37-114,37-1I5,37-116,37-I17, 37-117A,37-118,37-119,37-120,37-121,37-122,37-123,37-124,37-125,37-126,37-127,37-128,37-O O
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ras, a or a Degradaten t
Number Lines in Welds in Degradation Degradation Mechanism Consequence Risk Risk Segment ID ofWelds Segment Segment Mechanisms Group ID Category Category Category Rank 129,37-130,37-131,37-132 //37-077 t
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i Calculation No. A PENG-CALC-014, Rev. 00 Page D1 of D5 i
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APPENOlX D i
QUAllTY ASSURANCE VERIFICA TION FORMS 4
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i ABB Combustion Engineering Nuclear Operations
Calcul:ti:n No. A-PENG-cal.C-014 R:v. 00 Page D2 of DS Verification Plan
Title:
Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the CVCS at ANO 2 Document Number: C-PENG-CALC-014 Revision Number: 00 Instructions:
Describe the method (s) of verification to be employed, i.e., Design Review, Altemate Analysis, Qualification Testing, a combination of these or an attemative. He Design Analysis Verification 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 similar analyses, etc.
Descrintion of Verification Method-An independent review will be conducted as appropriate with the work activities described in Project Plan PP-2006839, Revision 00. The verification willinclude:
1.
Verification of a Design Analysis 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 format,... etc. are made in accordance with Design Analysis Verification checklist.
Verification Plan prepared by:
, Approved by: =
\\ ~
R. e. JA Qv r TH
- R ~@ Q.%0thaCCD:h0 T i independent RevN:wcr pnnied name andstenature
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Management approver pnhed name and signature
(
T ABB Combustion Engineering Nuclear Operations
C:Icul: tion No. A PENG CALC-014, Riv. 00 Page 03 of DS Other Design Document Checklist (Page1of3)
Instructions; The independent Reviewer is to complete this checklist for each Other Design Document. This Checklist is to be made part of the Quality Record package, although it need not be made a part of or distributed with the document itself. The second section of this checklist lists potential topics which could be ri evant for a particular"Other Design Document', if they are l
applicable, then the relevant section of the Design Analysis Verification Checklist shall be completed and attached to this checklist.
(Sections of the Design Analysis Verification Checklist which are not used may be left blank.)
Title:
Implementation of the EPRI Risk-Informed InsetTice Inspection Evaluation Procedure for the CVCS at ANO-2 Document Number:
Revision Number:
A-PENG-CALC-014 00 Section 1: To be completed for all Other Design Documents Yes N/A Overall Assessment 1
Are the results/ conclusions correct and appropriate for their intended use?
O 2
Are alllimitations on the results/ conclusions documented?
E Documentation Requirements 1.
Is the documentation legible, reproducible and in a fonn suitable for filing and retrieving as a Quality y
Record?
11.
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 S
O P
design inputs (in accordance with QP 3.2)
B O
C.
conclusions O
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 I.
Are all assumption identified, justified and documented?
8 O
II.
Are all assumptions that must be cleared listed?
O S
A.
Is a process in place which assures that those which are CENO responsibility will be cleared?
O O
B.
Is a process in place which assures that those which are the customer's responsibility to clear will O
B V
be indicated on transmittals to the customer?
ABB Combustion Engineering Nuclear Operations
C:Icul:ti:n No. A PENG CALC 014, lhv. 00 foge D4 of DS Other Design Document Checklist h
(Page 2 of 3)
Assessment of Signineant Desiga Changes Yes N/A I.
Have signincant design-related changes that might impact this document been considered?
g II.
If any such changes have been identined, have they been adequately addressed?
O O
Selection of Design inputs l.
Are the design inputs documented?
g Are the design inputs correctly selected and traceable to their sourcs?
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 specine to the information utilized?
g V.
Are the bases for selection of all design inputs documented?
g VI.
Is the verincation status of design inputs transmitted from customers appropriatYand documented?
8 O
Vil.
is the verification status of design inputs transmitted from ABB CENS appropriaiand documented?
O O
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.?
Peferences 1.
Are all references listed?
8 II.
Do the reference citations include sufTicient information to assure retrievability and unambiguous location of the referenced material?
Section 2: Other Potentially Applicable Topic Areas-use appropriate sections of the Design Analysis Verification Checklist (QP 3.4, Exhibu 3.4 5) and attach.
Yes N/A 1.
Use of Computer Software O
S 2.
Applicable Codes and Standards O
E 3.
Literature Searches and Background Data O
S 4.
Methods O
E 5.
Hand Calculations O
E 6.
List of Computer Software O
S.
7.
List of Microfiche O
O 8.
List of optical disks (CD-ROM)
O 8
9.
List of computer disks O
O
^
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i m ABB Combustion Engineering Nuclear Operations 1
Ccicul: tion No. A PENG CAI.C-014, R:v. 00 Page DS of DS C
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Other Design Document Checklist (Page 3 of 3)
Independent Reviewer's Comments Comment Reviewer's Comment
Response
Author's Response
Response
Number Required?
Accepted?
1.
Ednorialcomments seepages Yes Editorial changes have been Yes 3,5,6,7,8,11,14,15,16,17 made accordingly 18,20,21,22,31,33, A 3, A 8, A 9, A-10, C-0, C-1 2.
Para. 3.2.8 Does the LOCA Yes A review of the ANO 2 SAR Yes analysis credit chargingi is shows that charging is not there true redundancy between explicitly reguired for HPSI and charging?
mitigating a LOCA. Acre is no tme redundancy between HPSI and charging.
Hence,the paragraph has been edited for clarity.
3.
Why are the four check valves Yes Because the boric acid Yes shown closed?
makeup portion of the CVCS operates periodically,the check valves are shown in the closed position. De check valve symbolhas been h
changed to be consistent
'V with other reports.
4.
Table 3 is missmg consequence Yes Consequence CVCS-C-07 Yes CVCS-C-07. Is this a typo that has been added to Table 3.
may have introduced other its ommission did not errors in the table?
introduce any errors.
5, Figure 2 is missing CVCS-C-04 Yes Consequence CVCS-C-04 Yes has been added to Figure 2 6.
Need to resolve..:m on page 31 Yes item on page 3 I has been Yes resolved.
7.
Table I in Appendix A needs a Yes Page number has been added Yes page #
to Table 1 of Appendix A.
Checklist completed by:
Independent Reviewer ko8Ff 7- [, J/)pp/Tg jf
- M97 Prmted Name Signaturc
//
Date mU ABB Combustion Engineering Nuclear Operations
.