Text

Rev 0 to A-PENG-CALC-012, Implementation of EPRI Risk- Informed ISI Evaluation for RCS at ANO-2
	ML20217E929
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: 2CAN099706, Forwards Results of Pilot Plant Application Study Performed at Ano,Unit 2 for Review.Prompt Staff Review Is Requested. Formal Request for NRC Approval Will Be Made Following Submittal of Svc Water Sys Risk Evaluation in Dec 1997; ML20217E921; ML20217E929; ML20217E938; ML20217E950; ML20217E967; ML20217E987; ML20217F007; ML20217F029; ML20217F092;
References
	A-PENG-CALC-012, A-PENG-CALC-012-R00, A-PENG-CALC-12, A-PENG-CALC-12-R, NUDOCS 9710070309
	Download: ML20217E929 (132)
	v • d • e

_ _ _ _____________ ._

( Arkansas Nuclear One - Unit 2

Pilo: Plant Stuc.y s

i Risk-Informed Inservice Inspection Evaluation for the l Reactor Coolan1: System j

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A PENG CALC 012 R:v::lon 00 l' Design Analysls Title Page Page I cf 49

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Title:

Implementation of the EPRI Risk Informed Inservlee inspection Evaluation Procedure for the Reactor Coolant System at ANO 2 Document Number: A PENG CALC-012 Revision 00 Number:

Quality Class:

O QC.1(Safety Related) O QC 2 (Not Safety Related) O 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 Cogntrant Engineer (s) R. a. Weston h $/4I)

A. V. Bauer 6 g g Mentor g None

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Independent Reviewer (s) R. E. Jaquith [d/[gk g[$/fy

/ \ ['l' C] Management Approval D.T.Lubm QA Q k 8 j 0; .)

Project Manage N

2. Package Contents (this section may be completed after Management approval):

Total page count, including body, appendices, attachments, etc. 129 List associated CD-ROM disk Volume Numbers and path names: g None Note: CD-ROM are stared as separate Quality Records CD-ROM Volume Patn Names (to lowest directory which uniquely apphes to this document)

Numbers Total number of sheets of micronch:: 9 None Number of sheets:

Other attachments (specify):

3. Distribution:

B. Boya (2 copies)

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A PENG CALC 012 R:visi:n 00 Pag 2 2 cf 49 RECCRD OF REVISIONS

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Rev Date Pages Changed Prepared By Approved By 00 bb % Original A. V. Bauer R. E. Jaquith R. A. Weston O

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A bB l^\ Colculation No. A.PENG. CALC.012, Rev. 00 t

G) ~ Page 3 of 49 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE...............................................................................................................5 2.0 SC0PE..................................................................................................................5 3.0 SYSTEM IDENTiflCA TION AND BOUNDARY DEFINITION ....................................... .... 6 4.0 C0NMQ UENCE EVA L UA Tl0N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 4.1 C0NSEQ UENCE A SSUMP TlONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 4.2 CONSEQ UENCE IDEN TIFICA TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.3 SHUTDOWN OPERA TION AND EXTERNAL EVENTS........................................... 14 5.0 DEGRA DA TION MECHANISMS EVA L UA TION .. . ... ......... . .. . ... ......... . . . ..... .... . . . ...... . .. ... . 23 5.1 OA MA GE GR O UPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 5.2 DEGRADA TION MECHANISM CRITERIA AND IDENTIFICA TION........................... 24

5. 3 BA SIC DA TA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 6.0 SER VICE HIS TOR Y A ND SUSCEP TIBILI TY REVIEW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32

7. 0 RISK E VA L UA TlON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 8.0 EL EMENT SEL EC Tl0N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 0

9. 0 REFERENCES........................................................................................................47 LIST OF TABMS

() NUMB (R PAGE Li 1 R CS B 0 UNDA RIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2 RCS CONSEQ UENCE A SSESSMENT SUMMA R Y. . . .. . .... .... . . ... ...... .... ..... ...... .. ... . ..... ... . . 16 3 RCS CONSEQUENCES, FIGURES AND ISOMETRIC DRA WINGS................................... 18 4 DA MA G E GR O UPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5 DEGRADA TION MECHANISM CRITERIA AND SUSCEPTIBLE REGIONS......................... 25 6 SERVICE HISTORY AND SUSCEPTIBILITY REVIEW REACTOR COOLANT SYSTEM....... 34 7 RISK SEGMEN T IDEN TIFICA TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 6 8 RISK INSPEC TION SC0PE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 9 ELEMENT SELEC TION RISK CA TEG OR Y 4 .. . ........ ... .. . . . ...... . ....... .... ... .. . ... . .... . . .. ... .. .. 41 10 EL EMEN T SEL EC TION . RISK CA TEG O R Y 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 11 ELEMENT SELEC TION RISK CA TEGOR Y 5 .. . . .... . . . ... . . .. ... .. .... .. ............ .... .. . ...... . . . . .. .. 4 6 LIST O' lGURES NUMBER PAGE 1 REA CTOR COOLANT SYSTEM SIMPLIFIED SCHEMA TIC .............................................. 8 2 RCS HO T A ND COL D L EG FL O W PA THS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3 MA IN A ND A UXILIA R Y SPRA Y FL O W PA THS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 4 PRESSURIZER PRESSURE REllEF FL O W PA THS ... . . .... ... . .. .. . .... . . . .. .. . . .......... ... . ....... .... 21 5 RCS HO T AND COLD LEG DRAIN FL O W PA THS.. . .. . . . . .. . . .... . . ..... . . . ...... ......... . . .. .. .. . ..... 22 o

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( Calculation No. A-PENG-CALC-012, Rev. 00 i L) Page 5 of 49 1.0 PURPOSE The purpose of this eve'vatiots is to document the implementation of the Electric Power Research Institute (EPRI) Risk Informed Inservice Inspection Evaluation Procedure (RISI) of Reference 9.1 for the Reactor Coolant System (RCS) 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 approp;iate 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 Cunbustion Engineering Nuclear Operations Quality Procedures Manual (QPM-101).

2.0 SCOPE This evaluation applies to the RCS at ANO 2, and utilizes the ISIS Software (Reference 9.2),

which has been specifically developed to support and document the implementation of the EPRIRISIprocedure.

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

'a 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 canegory 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.

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%IFIF O Calculation No. A PENG CALC-012, Rev. 00 O Page 6 of 49 3.0 SYS1L?A IDENTIFICA TION AND BOUNDARY DEFINITION

3.1 System Description

The reactor is a pressurized water reactor with two coolant loops. The Reactor Coolant System (RCS) circulates water in a closed loop, removing heat from the reactor core and intemals and transferring it to the secondary (steam generat:ng) system. The steam generators provide the interface br* ween the reactor coolant (primary) system and the main steam (secondary) system. The RCS, by virtue of being a closed system, prevents the release of radioactive materials into the containment.

System pressure is controlled by the pressurizer, where steam and water are maintained in thermal equilibrium. Steam is formed by energizing immersion heaters in the pressurizer, or is condensed by the pressurizer spray to limit pressure ariations caused by contraction or expansion of the reactor coolant. The average temperature of the reactor coolant varies with power levels and the fluid expands or contracts, changing the pressurizer water level.

The RCS primary functions are:

To remove heat from the reactor core and internals and transfer it to the secondary system by the forced circulation of pressurized borated water which serves both as a e coolant and neutron moderator.

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To serve as a pressure boundary having a high degree nf leak tightness during normal

& operation. The integrity of this pressure boundary is asuured by appropriate recognition of operating, seismic and/or accident stress loading.

3.2 System Boundary The RCS is defined consistent with the FSAR (Reference 9.3). The scope of this analysis includes all Class 1 piping in this system which is currently examined in the ANO 2, ASME Section XI inservice Inspection flSI) Prt, gram (Reference 9.6). Lines which are part of the RCS were evaluated to determine their risk significance. The system boundaries are defined in Figure 1 and Table 1. Certain line segments contain welds that were not entered in the database (Reference 9.2) as outlined below:

3.2.1 Lines Downstream of the RCS Hot & Cold Leg Drain Valves (2HCC-1-2", 2HCC-2-2", 2HCC-3-2' 2HCC-4-2", 2HCC-5-2")

These line segments provide drain paths from the RCS hot and cold legs to the Reactor Drain Tank (RDT). These Ynes are normally isolated from the hot and cold legs by the manualdrain valves. A failure in any of these segments would not cause an initiating event if the valves are closed. If passive failure of both valves were to occur, the line segment would be exposed to normal operating condition. By exposing the segment to normal operating conditions, the potential for a small LOCA exists. Because both valves must fail, the resulting consequence would therefore be LOW. No degradation mechanisms were identified for the welds in these line segments. Because of the LOW consequence category and no damsge potential, the

,y risk significance of these segments would be LOW (i.e'., CAT 7). Since no element U

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l A R Ik E'% IF BF Calculation No. A-PENG-CALC 012, Rev. 00 l

. Page 7 of 49 selections are needed for low risk significant segments, the welds for these lines were not ontered in the database.

3.2.2 Lines Downstream of Pressurizer Safety Relief Valves & Downstream of Pressurizer Vent /LTOP Valves (2FCC 16', 2FCC-1 10', 2FCC 2 3', 2FCC 2 3')

These lines provide pressure relief paths from the pressurizer to the quench tank.

The pressuriter safety valves protect the RCS from over-pressurization resulting from design basis transients that cause an increase in RCS pressure. Pressurizer pressure relief is also provided by the ECCS or LTOP vent valves b the event that secondary side heat removal is lost (i.e., total loss of feedwater to the steam generators). Pressurizer pressure relief via the "%ty valves, ECCS or LTOP vent valves results in a rupture of the quench tans. ~,)ture disk which is of no safety significance. Because the segments are isola.ud, a failure would not cause an initiating event. A failure in any of the above line segments would also have no safety significance, since these lines are not needed to support or accomplish any of the safety functions following a design basis event. This is true for any line that interfaces with the quench tank. Because the above line segments are not needed to support any safety function, a LOW consequence category is assigned. Based on this assigned consequence category, the risk significance of the 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 Une from RCS Cold leg to Letdown Isolation Valve (2CCA-12-2")

This line provides RCS letdown for chemistry control. This line segment is included as part of the Chemical and Volume Control System and is therefore not evaluated

, as part of the RCS.

3.2.4 Lines with Nominal Diameter of I' or Less Piping with a nominal diameter of I' or less was not explicitly evaluated to determine its risk significance. Since volumetric examination of this piping is not practicable, the most 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 pipinry is not warranted.

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C8/Cul8 tion NO. A-PENG-CALC-012, Rev. 00 P8ge 9 Of 49 TABLE 1 RCS BOUNDARIES Lme Lme Desceptoon ISI Pye Nomonel Pte Pro Thickness (m)

Number Drawing Cless Diameter (m)

Number 2BCA 112* 1+essunzer surge hne D-6310601-023 1 12 1.126 2BCA-14-3

Pressunzer vent bwass hne 2BCA 14-1 1 3 0.438 2BCA 14-4* Pressurizer vent kne - 4

section 2BCA 14-1 1 4 0.438 2BCA.166* Pressurizer vent hne 2BCA 14-1 1 6 0.662 2CCA 1-42* Hot leg from reactor vessel to steem D-6370601-001 1 42 3.75 generator 2E-24A 2CCA 1030' CoM leg from reactor coolant purre C-6370601 109 1 30 3 2%32D to reactor vessel 2CCA 13 3* Pressunter main sprey kne from RCS 2CCA 13-1 1 3 0.438 coM leg 2W32A 2CCA 14-3* Pressurder mom sprey kne from RCS 2CCA 14-1 1 3 0.438 coM leg 2F32B 2CCA 16 3* Pressunter mem sprey kne 2CCA 161,2CCA. 1 3 0.438 16-2, 2CCA 16-4 2CCA 16-4

Pressurder mem sprey 4* kne 2CCA 161,2CCA. 1 4 0.438 16-2, 2CCA 16 4 2CCA 16-2

Auxihery sprey kne from motor valve 2CCA 16-1 1 2 0.344 2CV-48242 to common sprayline 2CCA 2-42

Hot leg from reactor vessel to steem C-AAK 611006 1 42 3.75 generator 2E-248 2CCA 29-2* RCS coM leg 2%328 drein line 2CCA 291 1 2 0.344 2CCA 3 30* CoM leg from steem generator 2E-24A C-ARK 611-009 1 30 3 to reactor coolant pump 2W32A 2CCA 362

RCS coM leg 2P 32C dreon kne 2CCA 301 1 2 0.344 2CCA 312* RCS coM100 2F320 drem kne 2CCA 31 1 1 2 0.344 2CCA 32 2* RCS hot leg to steem generator 2E- 2CCA 32-1 1 2 0.344 24A drain hne 2CCA-4-30* CoM leg from reactor coolant pump C-ARK-611-010 1 30 3 2F32A to reactor vessel 2CCA-6-30* CoMleg from steem generator 2E 24A C-ARK 6 91007 1 30 3 to reactor coolant pump 2W328 2CCA 6 30' CoMleg from reactor coolantpump C-ARK 611008 1 30 3 2P-32R to resetor vessel 2CCA.7-30' CoMleg trorr steem generator 2E-248 C ARK 611011 1 30 3 to reeetor coolant pump 2F32C 2CCA-8-30* CoM leg from reactor coolant pump C-ARK 611012 1 30 3 2F32C to reactor vessel 2CCA 9-30* > CoM leg from steem generator 2E 248 C-6370601 108 1 30 3 to reactor coolant pump 2W32D 2BCA 0-6* Pressunzer Safety Velves (Lme Not Avedeble 1 6 0.662 number is assigned to account for two welds associated with PSVs O

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4. 0 CONSEQUENCE EVALUA TION The Reactor Coolant System (RCS) circulates borated water through the reactor core in a closed loop. This circulation removes heat from the core and reactor internals and transfers the heat to the steam generators to produce steam for the secondary system. The reactor coolant is circulated through the core by the four centrifugal reactor coolant pumps. In addition to the transfer of heat, the RCS serves as a neutron moderator and provides one of the three barricts that prevent the release of fission products to the atmosphere. The RCS has a design pressure of 2500 psia and a design temperature of 6507 (except the pressurizer which has a design temperature of 7007). At normalpower operating condition (i.e.,100% power), approximately 120 x 10* Ib/hr of reactor coolant is circulated through the two closed loops connected in paraHel to the reactor vessel. The RCS is interconnected with other plant systems to provide makeup and reactivity control. During normal power

, operations, the Chemical and Volume Control System (CVCS) provides the makeup for l inventory lost, and maintains coolant chemistry and reactivity control. During a loss of Coolant Accident (LOCA), the Emergency Core Cooling System (ECCS) provides the necessary makeup and core cooling functions. The consequence evaluation of these systems are treated separately, and are therefore not includedin the consequence evaluation of the RCS.

The consequence evaluation for the RCS was performed based on the guidance provided in the EPRI procedure (Reference 91). The evaluation focused on the impact of each pipe segment failure on the capability of the RCS to perform its design functions, and on the

[3 overaH operation of it.e plant. Impacts due to direct and indirect effects were considered.

& GeneraHy, the direct effects are LOCA initiating events. An indirect effect resulting from the failure of a pipe segment would affect neighboring equipment within the RCS or other interfacing system (s). Indirect ehects 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 RCS is located inside the containment. Certain safety related components which are designed to mitigate a LOCA are also locatedinside the containmerat.

In general, the spatial effects of segment failure are primarily associated with flooding, spraying, or jet impingement. Dynamic analyses (Reference 9.3, Section 3.6.4.2) have been performed to assess the spatial effects of failed RCS piping. The.,e analyses have concluded that a failure in the RCS piping wiH not cause a more severe 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 affected. However, the ability of the engineered safety features to mitigate the consequences of a LOCA wiH not be impaired.

The environmental effects caused by a failure of the RCS piping have also been assessed.

AH safety injection components located inside the containment have been designed to withstand the LOCA environment (Reference 9.3, Section 6.3.2.12.1). The containment cooling units are also designed to maintain their functional integrity in a post LOCA environment (Reference 9.3, Section 6.2.2.2.2).

A system walkdown is usuaHy conducted as part of the consequence evaluation. Because

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, the RCS is located inside the containment, access to the containment would be required to (O conduct a walkdown of this system. At the time this evaluation was being performed ANO-ABB Combustion Engineering Nuclear Operations

A Ik R MIF19 Calculation No. A PENG-CALC 012, Rev. 00 Page 11 of 49 2 was operating at normalpower. Access to the containment was therefore not feasible.

The dynamic analyses of a failure in the CS piping and the assessment of engineered safety features to withstand the post-LOCA environment have concluded that the mitigation of a LOCA win not be impaired. The dynamic analyses and the post-LOCA environment assessment have addressed the spatial effects of an RCS pipe failure. It is therefore believed that a walkdown of the RCS wouldprovide no new insights regarding the impact of spatial effects andis not needed.

Assumptions used in the consequence evaluation are discussed in Section 4.1. Fifteen consequence segments were identified for the RCS lines entered in the database. Of ,the fifteen, nine were assigned as HIGH and six as ' MEDIUM'. The consequence assessment summary for these segments is provided in Section 4.2. The bases and justifications for each category assignment are provided in Appendix A. This appendix contains reports obtained from the ISIS software (Reference 9.2) for the RCS. For the RCS lines not entered in the database, tv o were assigned as LOW consequence (see Section 3.2).

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 rondemness are piping segments which are not routinely exposed to the normal operating pressures of the RCS. In these cases, it is assumed that the piping is so weakened by the degradation process that upon demand there would be a sudden failure. For portions of the RCS where the piping is normany or regularly exposed to operating pressures, it is assumed that the piping failure is most likely to cause a design basis LOCA. SpecificaHy, it is assumed that:

(a) Piping that is within the RCS cold legs can only failin such a way as to cause a large LOCA initiating event and also to cause the unavailability of one of the fourinjection loops to the RCS.

(b) Piping that is downstream of the normaHy closed drain valves is usuaHy not exposed to RCS operating pressures, and is therefore assumed to failif a sman LOCA potential exists.

(c) Piping that is between the normaHy closed LTOP/ECCS vent valves is not routinely exposed to RCS operating pressures. For this piping, it is assumed that the fai'nre occurs during a demand or if the potential for exposing the segment to RCS operating conditions exists.

d) Piping that is upstream of the auxiliary spray line check valve and downstream of the auxiliary spray line isolation valve does not routinely experience the fuH discharge pressure of the charging pumps. For this piping, it is assumed that the failure occurs during a demand or if the

potentialfor exposing the segment to RCS operating conditions exists.

4.1.2 RCS pipe segment failures that cause a LOCA are assumed to occur during the fuH power operating condition of the plant. .

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'G) Page 12 of 49 4.1.3 A segment failure in the RCS has the potential to (1) lead to a LOCA that requires a protective response from the Reactor Protection System and/or the Engineered Safety Features, (2) degrade the reliability of the RPS or the Engineered Safety Features, or (3) both 1 and 2. Specifically, it is assumed that:

(a) Based on the Individual Plant Examination for ANO 2 (Reference 9.16 Tables 3.11&3.13), a large LOCA is assumed for segment failures that occur in RCS piping of nominal diameter greater than 4'. A medium LOCA is assumed for segment failures that occur in RCS piping of nominal diameter between 2* .:nd 4'. Segment failures occurring in RCS piping of norninal diameterless than 2* are assumed to cause a smallLOCA. In addition to the nominal diameter, the failed piping segment must also be exposed to normal fullpower operating pressures of the RCS in order for a LOCA to occur.

(b) For pipe segment failures that require a mitigating system demand, the potential for break isolation is evaluated and if the break is readily detectable and isolable, then isolation capability is treated as one additional success path, or the equivalent of having one additional train of redundancy available.

If the break isolabilitv is less apparent, then the unisolated configuration is used to determine th.' consequence category.

4.1.4 A failure in the auxiliary spray line is assumed to be detectable because of the close observation of the pressurizer tcmperature during auxiliary spray operation. This observation is needed to satisfy the requirements of Table 5.71 of the Technical

) Specifications (Reference 9.14), i.e., to evaluate the pressurizer spray nozzle fatigue caused by the dT.

4.1. 5 Based on the dynamic analyses of RCS pipe breaks and the capability of mitigating components to withstand a LOCA environment, it is assumed that spatial effects of a segment failure in the RCS piping are negligible and of no concem.

4.1.6 The main pressurizer spray valves are configured to operate in the automatic mode.

4.1.7 A potential LOCA initiating event was considered for pipe segments normally isolated from the RCS (i.e., auxiliary spray, ECCS/LTOP and drain lines). A potential LOCA is defined as the failure of the isolating valve followed by the failure of the associated RCS segment. The resulting consequence for a potential LOCA is based on the failure potential of the isolating valve and the Conditional Core Damage Probability (CCDP) for the type of LOCA. Specifically, la) The nominal diameter of the auxiliary spray segment is 2', thus the potential initiating event in this segment is a small LOCA. The probability of the check valve failing to maintain the segment isolated during normalpower operatio 1 is approximately 1.0E-02 (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 MEDIUM consequence.

(b) The potentialinitiating event for the drain line segment is also a small LOCA

, because the nominal diameter of the drain pipe is 2". The failure probability

\

J 6

of the manual valve is approximately 4.0E-2 (Reference 9.16). The ABB Combustion Engineering Nuclear Operations

i ABB Calculation No. A.PENG. CALC-012, Rev. 00 Page 13 of 49 combined effect of the failure of the manual valve and the CCDP for small LOCA (Table 1 of Appendix A) results in a MEDIUM consequence.

(c) The nominal diameter of the LTOP/ECCS segment is between 3' and 4',

thus the potentialinitiating event in this segment is a medium LOCA. The probability of the motor-operated valve failing to maintain the segment isolated during normalpower operation is approximately 3.0E-03 (Reference 9.16). The combined effect of the motor-operated valve failure and the CCDP for medium LOCA (Table 1 of Appendix A) also results in a MEDIUM consequence.

(d) By including the failure potential of the second isolation valve in series for any of the above 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.

Based on the failure potential of the isolating valve and CCDP for the associated LOCA, the resulting consequence is MEDIUM for normally isolated segments with single isolation. For segments with double isolation, the resulting consequence is LOW.

4.2 CONSEQUENCE IDENTIFICATION The consequence summary assessment is provided in tabular form in this section. Simplified ,n schematics are provided in Figures 2 through 5 to illustrate the boundaries for each of the J RCS consequences. Dotted lines are used to identity the boundaries for each consequence.

Major RCS equipment is shown on these figures for ease of identification. Table 2 summarizes the consequence evaluation for the RCS. This table contains the following information for each of the consequences identified:

Consequence ID 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 RCS, 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 provided in the EPRI procedure (Reference 9.1) t O

ABB Combustion Engineering Nuclear Operations

A It Ik

1IFIp (3 Calculation No. A-PENG. CALC-012 Rev. 00

.Page 14 of 49 The bases and Justifications for each of the assigned consequences are documented in Appendix A. The ISIS (Reference 9.2) software was used as a tool to prepare the documentation in this appendix. The documentation of the spatial effects are based on a review of the dynamic analyses provided in the ANO-2 Safety Analysis Report (Reference

9. 3).

4.3 SHUTDOWN OPERA TION AND EXTERNAL EVENTS Shutdown Operation The consequence evaluation is an assessment assuming the plant is at power. Generally, the al-powerplant configuration is cornidered 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 stable state. At power, the plant is critical, and is at higher pressure and temperature in comparison to shutdown operation. The current version of the methodology (Ref~ence 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.

Pipe segments that are already ranked as "HIGH' consequences from the evaluation at.

power need not be evaluated for shutdown. Those that are already ' MEDIUM" require some confidence that 'HIGH" would not occur due to shutdown configurations. However, the

LOW" consequences for power operation requires more confidence that a *HIGH'

( would not occur and some confidence that a ' MEDIUM" consequence would not occur.

G Recognizing this, a review and comparison of system consequence results for power operation versus potential consequence during shutdown operation was conducted.

The results of the comparison indicate that during shutdown the RCS operating conditions

. are less severe than during power operation. It was also noted that the consequence ranking for major piping is already 'HIGH" during power operation, thus enveloping the consequence ranking during shutdown.

Extemal Events Although extemal events are not addressed in the current version of the methodology (Reference 9.1), the potentialimportance of piping failures during external events is also considered. The ANO 2 IPEEE was reviewed to determine whether extemal initiating events, with their potential common cause impacts on mitigating systems, could affect consequence ranking. This information, along with information from other extemal event PRAs, is considered to derive insights and confidence that consequence ranking is not more significant during an extemal event. The following summarizes the review for each of the major hazards (seismic, fire and others).

Seismic ChaHenges The potential effects of seismic initiating events on consequence ranking is assessed by considering the frequency of challenging plant mitigating systems and the potentialimpact on the existing consequence ranking. The following summarizes this assessment:

Generally, the RCS piping considered in this evaluation has a seismic fragility p capacity much greater than the 0.3g screening value and is not considered likely to fail during a seismic event.

(U)

ABB Combustion Engineering Nuclear Operations

A It R 7%IFIF Calculation No. A PENG CALC 012, Rev. 00 Page 15 of 49

Failure of RCS piping during power operation already causes an initiating event. The frequency of an earthquake induced pipe failure in the RCS is less than the at power value assumed in the evaluation. Also, the likelihood of a simultaneous seismic event during or after a pipe break is low.

Reactivity controlis unlikely to be affected by seismic events because loss of offsite power (the most likely scenario) will de-energize and drop control rods. A very large earthquake may cause 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 brelevant ut this point and it is an extremely low probability event.

Based on the above, the consequence ranking for the RCS during a seismic event is enveloped by the at power consequence ranking.

Fire ChaHenges . The ANO-2 IPEEE 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 at-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 RCS.

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D'1 F EF Calculation NO. A-PENG-CALC-012, Rev. 00 Page 16 Of 49 Table 2 RCS Consequence Assessment Summary Consequence D Boundary Description DB Event Direct Effects Special Effects kaipact Grene AveneNe Midgeting Consequence Category Trains Ceregory RCS-C-O f Rgure 2 Loss of reactor coolant via N large LOCA initiating None Hristing event AR trains of the HIGH either RCS hot leg event. mitigering aystems RCS-C-02 Rgure 2 Loss of reactor coolant via W large LOCA initiating None Htiating event Tivee ECCS iriection HIGH RCS coldleg 2P-32A event and degredation kops of multiple systems RCS-C-03 Rgure 2 loss of reactor coolant via N large LOCA intriering None Hristang event T! vee ECCS Hection H:GH RCS coldleg 2P-328 ovent and degredation hops of multiple systems RCS-C-04 Rgure 2 loss o*.eector coolant via N large LOCA initiating None Hristing event T! vee ECCS ?petion HIGH RCS coldleg 2P-32C event and degrrdet'on hops of multiple systems RCS-C-05 Rgure 2 Loss of reactor coolant via W large LOCA htisting None Htiating event T! vee ECCS hiection HIGH RCS coldleg 2P-320 event and degredation hops of multiple systems RCS-C-OS Ryure 3 Loss of reactor coolant via IV Mediurn LOCA None Wriefing event AR treins of the HIGH main pressunzer sprey 6ne initiating event ^ end hss ef cne mitiget%g systems system RCS-C-07 Rgure 3 loss of reactor coolant via N Smen LOC 4 Mtiating None Htiating event AR treiru of the HKiH euxiliary sprey Ene event and kss of one mitigering systems sys ten

RCS-C-OS Rgure 4 Loss of reactor cootent vie IV large LOC 4 initiating None haitiating event AR treins of the HIGH pressurizer ventATOPEne event entigering systems RCS-C-09 Rgure 5 loss of reactor coolant via N Smen LOCA htiethg None Hristing event AR tret.ns of the HIGH RCS hot or cold tog drein Ene event mitigering systems RCS-C- 10 Rgure 5 1%tentialhss of reactor N Pbtential smeR LOC 4 Inkne None AR treins of mitigerm MEDIUM coolant vie hot leg drain Ene systems RCS-C-11 Rgure 5 lbtentien hss of reactor N Pbtential smeR LOC 4 None None AR trains of mitigsting MEDIUM cooient vie coldleg drain Ene systems 2P328 ABB Combustion Engineering Nuclear Operations

TP ARR 9% WlF Calculation No. A-PENG-CALC-012, Rev. 00 Page 17 of 49 Table 2 (Cont'd)

RCS Consequence Assessment Summary DB Event Direct Effects Spatial Effects kvect Grom AveWeblo Mitigering Coneequence Consequenew D Bourufery Descrktion Category Treine Ceregory P6tentief hss of reactor IV Ntontialsmet LOCA None None AB trains of mitigating MEDIUM RCS-C-12 Roure 5 coolant vie cohileg drain Gne systems 2P32C Rgure 5 1%tentistloss of reactor IV hrentief smeR LOCA None None AR e,au of mitigating MEDIUM RCS-C-13 coolent wie coldleg r* rein Ene systems 2P32D 1%tentiel smen LOCA None None AM tidu of mitigering MEDIUM RCS-C-14 Rgure 3 1%tentief kss of reacto? IV coolant wie muniNavy spreyline systems.

1%tentialloss of reactor P6tentief mediurn None None- AR eidu of the MEDIUM RCS-C- 15 Rgure 4 IV coolant via LTOP/ECCS vent LOCA mitigering systems Kne ABB Combustion Engin ing Nuclear Operations

A Ik R MIFBF

('} Calculation No. A PENG CALC 012, Rev. 00 G Page 18 of 49 Table 3 RCS Consequences, Figures and Isometric Drawings Consequence ID figure Isometric Drawings Number RCS-C 01 2 D-6370-501-001 C ARK 511-006 D-6370-501-023 RCS-C-02 2 C-ARK 511-009 C ARK 511-010 RCS-C-03 2 C-ARK 511-007 C-ARK 511008 RCS-C-04 2 C-ARK 511011 C ARK-511-012 RCS-C-05 2 C-6370-501 108 C-6370-501 109 RCS C-06 3 2CCA 131 1 2CCA 15-41 2CCA 15-1 1 2CCA.14 1 1 2CCA 15 21

_ RCS C-07 3 2CCA 161 1 2CCA 16-21 RCS-C-08 4 2BCA-14 1 1 RCS-C-09 5 2CCA 291 1 2CCA 30-1 1 2CCA 31 1 2CCA 32-1 1 RCS-C-10 5 2CCA 321 1 RCS-C-11 -

5 2CCA 291 1 RCS-C-12 5 2CCA 30-1 1 RCS C-13 5 2CCA 31 1 RCS C-14 3 2CCA-16 1 1 RCS C 15 4 2BCA 14-1 1

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/ h ABB Combustion Engineering Nuclear Operations

z pp Calculation No. A-PENG. CALC-012, Rev. 00 Page 19 of 49 PRESSURIZER

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TO QUENCH TO QUENCH TO QUENCH TANK TANK TANK d i J L , i ,

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~h Calculation No. A PENG. CALC 012, Rev. 00 (Y Page 23 of 49

5. 0 DEGRADA TION MECHANISMS EVALUATION 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 RCS system, as described in Section 3.2 of this report. The conditions considered in this evaluation are: design characteristics, fabrication practices, operating conditions, and service experience. The degradation mechanisms to be identified (Reference 9.1) are:

ThermalFatigue (TF)

Thermal Stratification, Cycling, and Striping (TASCS)

- Thermal Transients (TT)

Stress Corrosion Cracking (SCC)

- Intergranular Stress Corrosion Cracking (IGSCC)

Transgranular Stress Verosion Cracking (TGSCC)

- External Chloride Stre Corrosion Cracking (ECSCC)

- Primary Water Stress Cerrosion Cracking (PWSCC)

Localized Corrosion (LC)

- Microbiologically influenced Corrosion (MIC)

Pitting (PIT)

- Crevice Corrosion (CC)

Flow Sensitive (FS)

- Erosion Cavitation (E C)

(n Flow Accelerated Corrosion (FAC)

)

In performing this evaluation, some basic inputs were used. These inputs are discussed in Section 5.3. The criteria and justifications are provided in Section S.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 RCS T and RCS-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 each element are presentedin Appendix B.

Table 4 Damage Grou;~s Damage we Mechaniana Feiwe Group ThermalFetigue stron ' . i , Ges cking LoceEred Corrosion Row senekke Potential 20 TASCS TT IGsCC 'ou ECsCC PWsCC MIC PfT CC E-C FAC Cateeory RCs T Yes Yes No No No No No No No No No Smen leek RCS-N No No l No No llo No No No No No No None O

G ABB Combustion Engineering Nuclear Operations

ABB e ^

Calculation No. A.PENG CALC-0.'2, Rev. 30 Pare 24 of 49 5.1 DAMAGE GROUPS 5.1.1 DM GROUP: RCS4 and RCS-N The RCS-T damage group is conWdered subject to tioermal fatigue due te the potent,'st for thermal stratification (TASCS) an'J/or thermal transients (TT). The affected sections are described in the table below and depicted in Figures 2 5.

Une Number Affected Sections ThermalFatigue Cause 2BCA 112' Pressurizer Surge Une entire line from thermal stratification (horizontal section) hot leg surge line nozzle to pressurizer and thermal transients (entire line) surge nozzle 2BCA 14-6' Pressurizer LTOP Piping entire section thermalstratification from pressurizer LTOP nozzle connection to both 6' x 4' reducers (excludes upper verticalportion of 6' tee) 2BCA 14-4' Pressurizer LTOP Piping both sections thermal stratification lentire section) and from 6' x 4' reducers to isolannm valves therraal transients tblock valves 2CV.

2CV-47312 and 2CV-4741 1 473G-1 and 2CV-4740-2 to isolation valves 2l.V-47312 and 2CV-4741-1) 2BCA 14-3' Pressuriser LTOP Piping entire section thermaitranstsnts from 4' x 4' x 3' reducing tee to isolation vai.s 2CV-46981 2CCA 154' Pressurizer Main Spray Poping entire thermal stratification (horizontal section section from auxiliary spray tee upstream of pressurizer spray nozzle) and connection to pressurizer spray nozzle thermal transients (entire section) 2CCA 16 2' Pressurizer Auxiliary Spray Piping - thermal stratification and thermal horizontal section from tee connection transients back upstrearn to first elbow The RCS-N damage group is not considered susceptible to any damage mechanism, and includes the hot and ccid legs, drain lines, main spray piping upstream of the auxiliary sp ay tee connection, and the auxiliary spray piping located upstreem of the first elbow located upstream of the tee connection into the main spray line (See Figures 2-5).

S.2 DEGRADATION MECHANISM CRITERIA AND IDENTIFICAT!ON The degradction mechanisms and criteria assessed are presented in Table 5.

T l

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N !k !I, F%IF F

/% Calculation No. A PENG CALC 012, Rev. 00

'Page 25 of 49 Table 5 Degradation Mechanism Criteria and Susceptible Regions Degradation Mechanism Crkala Suscep@c Redons TF TASCS -nps > 1 inch, and non!es, branch pipe

-pipe segment has a slope < 43*from hort:ontal (includes elbow or connections, safe ends, tee into a verticalpipe), and welds, heat afected zones

-potential existsfor lowpow in a pipe section connected to a II 'b"*#I*""U component allowing mixing ofhot and coldpuids. or potential existsfor leakagepow past a valve (i.e., in-leakage, out-j

" 'f' '""

leakage, cross-leakage) allowing mixing ofhot and coldpuids, or potential exhtsfor convection heating in dead-endedpipe s ctions connected to a source ofhotpuid, or potential existsfor two phase (steam / water) pow, or potential existsfor turbulentpenetration in branch pipe connected to headerpiping containing hotpuid with high turbulentpow, and

-calculated or measured AT > $0*F, and

-Richardson number > 4.0 TT -operating temperature > 270*Ffor stamiess steel, or operating temperature > 220*Ffor carbon steel, and p -potentialfor relatively rapid temperature changes includmg coldpuidinjection into hotpipe segment, or hotpuid injection into coldpipe segment, and

- l JT > 200*Ffor stainless steel, or lAT > 150*Ffor carbon steel, or l AT > ATallowable (applicable to both stainless andcarbon)

SCC IGSCC -evaluated in accordance with existing plant IGSCCprogram per austenitic stainTss steel (BWR) NRC Generic Letter 88-01 welds andflAZ IGSCC -operating temperature > 200*F, and (PWR) -susceptible material (carbon content 2 0.033%), and

-tensile stress (includmg residual stress) is present, and

-oxygen or oxidaing species are present OR

-operating temperature < 200*F, the attributes above apply, and

-initiating contaminants (e.g., thiosulfate,fuoride, chloride) are also required to be present TGSCC -operating temperature > iS0*F, and austenitic stainless steel

-tensile stress (including residual stress) is present, and base metal, welds, and

-halides (e.g.,fuoride, chloride) are present, or H4Z caustic (NaOH)is present, and

-oxygen or oxidi:ing species are present (only required to be present in conjunction w/ halides, not required w/ caustic)

O Q,0 ABB Combustion Engineering Nuclear Operations

A R Ik MIFID Calculation No, A PENG-CALC-012, Rev. 00 Page 26 of 49 Table 3 (cont'd)

Degradation MechanLsm Criteria and Suscentib'e Regions

", , Criteria Susceptible Regions SCC ECSCC -operating tempe~ature > 150*F, and austenitic stainless steel

-tensile stress is present, and base metal, welds, and

-an outside piping surface is withinfive diameters ofa probable llAZ leak path (e.g., valve stems) and is covered with non-metallic ins *lation that is not in compliance with Reg. Guide 1.36, or ,,

an outside piping surface is exposed to wettingfrom chloride bearing environments (e.g., seawater, brackish water, brine)

PWSCC -piping materialisinconel(Alloy 600), and no::les, welds, andHAZ

-exposed to primary water at T > 620*F, and without stress relief

-the material is mill annealed and cold worked, or cold worked and welded without stress relief LC MIC -operating temperature < 150*F and fittings, welds. HAZ,

-low or intermittentflow, and base metal, d:ssimilar

-pH < 10, and metaljoints (e.g., welds,

-presenceantrusion oforganic material (e.g., raw water system), or flanges), and regions water source is not treated w4iocides (e.g., refueling water tank) containing crevices PIT -potential existsfor lowflow, and

-oxygen or oxidi:ing species are present, and

-initiating contaminants (e.g., fluoride, chloride) are present CC -crevice condition exists (e.g., thermal sleeves), and

-operating temperature > 130*F, and

-oxsgen or oxids:ing species are present FS E-C -operating temperature < 250*F, and fittsngs, welds, HA, and

-flow present > 100 hrs)r, and base metal

-velocity > 30ft's, and

-(Ps - PJ / AP < 3 FAC -evaluated in accordance with existing plant FA Cprogram per plant FACprogram O

ABB Combustion Engineering Nuclear Operations

A R Ik MIF ER 0 Calculation No. A-PENG-CALC-012, Rev. 00 Page 27 of 49 5.2.1 ThermalFatigue (TF)

Thermal fatigue is a mechanism caused by attemating stresses due to thermal cycling of a component which results in accumulated fatigue usage and can lead to crack initiation and growth.

5.2.1.1 ThermalStratification, Cycling, and Striping (TASCS)

The applicable piping sections from the RCS-T damage group, as defined in Section 5.1, are considered subject to thermalstratification.

ANO-2 plant-specific thermocouple data does not exist for the pressurizer surge line (2BCA 1-12"). However, CEOG Task 662 (Reference 9.18), which discusses instrumentation of the surge line at three CE plants, indicated large top-to-bottom temperature differentials caused by either an insurge or outsurge of the pressurizer.

Additionally, NRC Bulletin 8811 states that surge line thermal stratification is a generic issue for all PWR's. These fluid surges can result from various plant heatup or cooldown operations (e.g., spray initiation, energizing heaters, mismatch of charging and/or letdown flow). Therefore, the potential exists for thermal stratification in the pressurizer surge line in all PNR's.

CEOG Task 827 (Reference 9.19) assessed the potential for thermal stratification in Pressurizer Safety Valve (PSV) and Low Temperature Over-Pressurization (LTOP) piping (2BCA 14-6' and 2BCA 14-4'). ANO-2 thermocouple data, recorded during v an entire fuel cycle, indicated the presence of ATs of between 100*F and 180*F during plant heatups and cooldowns. These top-to-bottom temperature differentials are believed to be the result of two phase stratification (i.e., presence of steam and cooler condensate). Although thermocouple data does not exist for the portions of LTOP piping line 2BCA 14-4' downstream of the normally closed block valves, these horizontal sections are potentially subjected to thermal stratification during plant cooldowns when the block valves are opened to place the LTOP relief valves in service and during LTOP relief valve blowdown transients. The potential for two phase flow is considered remote and will be dependent upon the quantity of condensate remaining in this piping after plant hectup.

ANO-2 plant-specific thermocouple data does not exist for the pressurizer spray line (2CCA 15-4'). However, CEOG Task 482 (Reference 9.22), describes another CE plant, which discovered large top-to-bottom temperature differentials due to two phase flow (i.e., steam over water) in the upper horizontal run of the main spray piping. This is a result of periods of low spray flow allowing steam from the pressurizer to backfillinto the main spray piping. Therefore, the potential exists for thermal stratification in the main spray piping lines in CE plants with similar spray line configurations.

CEOG Task 886 (Reference 9.20) assessed the potential for thermal stratification in auxiliary spray piping (2CCA 16-2'). Thermocouple data, recorded during heatup and cooldown transients, indicated the presence of ATs of up to 100*F. This is likely due to some minor valve leakage in conjunction with manipulations in charging f^ pump operations.

(

ABB Cornbustion Engineering Nuclear Operations

F% WIP Calculation No. A-PENG-CALC 012, Rev. 00 Page 28 of 49 5 2.1.2 Thermal Transients (TT)

The applicable piping sections from the RCS T damage group, as defined in Section 5.1, are considered subject to thermal transients.

Tne pressurizer surge line (2BCA 112") is subjected to thermal transients due to insurges and outsurges from the pressurizer with ATs occurring as high as 350*F (Reference 9.18). These are the same transients that cause thermal stratification in the horizontalrun of the surge line.

The sections of Low Temperature Over-Pressurization piping (2BCA 14-4' and 2BCA.14-3') downstream of the normally closed block valves are potentially subjected to thermal transients during pressurizer cooldown and LTOP Relief Valve Blowdown transients (Reference 9.21, Table 6.2 2), with an initial temperature of 120*F and a final temperature of 420*F.

The section of pressurize,. main spray piping (2CCA 15-4') downstream of the auxiliary spray tee connection is subjected to thermal transients during normalplant cooldown upon initiation of auxiliary spray flow, with an initial temperature of 355*F and a final temperature of 105*F (Reference 9.4, Figura 5).

The section of auxiliary spray piping (2CCA 16-2') extending from the tee connection to the main spray line back upstream to the first elbow is subjected to thermal transients during normal plant cooldown upon initiation of auxiliary spray flo w, with an initial temperature of 355*F and a final temperature of 105*F (Reference 9.4, Figure 5).

5.2.2 Stress Corrosion Cracking (SCC)

The electrochemical reaction caused by a corrosive or oxygenated media within a ~

piping system can lead to cracking when combined with other factors such as a susceptible material, 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 corrosion cracking.

5.2.2.1/ntergranular Stress Corrosion Cracking (IGSCC)

All of the piping in this system, with the exception of the auxiliary spray line, operates in excess of the IGSCC temperature threshold of 200*F. However, none of this piping is exposed to oxygen or oxidizing species (RCS backpressure maintains isolation valves closed preventing oxygenated RWT water supply intrusion) and is therofore not considered susceptible to IGSCC. The auxiliary spray line, which operates at less than 200*F (140*), only functions during plant cooldowns when the Reactor Coolant Pumps are removed from service. The auxiliary spray water supply is provided by the Volume Control Tank via the Chemical and Volume Control system which is maintained essentiaoy oxygen free. Consequently, this piping is also not considered susceptible to IGSCC.

O ABB Combustion Engineering Nuclear Operations

ARR MIFIP f3 Calculation No. A PENG-CALC-012, Rev. 00 O Page 29 of 49 5.2.2.2 Transgranular Stress Corrosion Cracking (TGSCC)

Plant chemistry controls ensure that the levels of halides or caustics present in the system are maintained extremely low and this piping is therefore not considered susceptible to TGSCC.

5.2.2.3 External Chloride Stress Corrosion Cracking (ECSCC)

ANO-2 complies with the requirements of Regulatory Guide 1.36 for non-metaHic thermalinsulation and conseqaently the potential for ECSCC to occur does not exist.

5.2.2.4 Primary Water Stress Corrosion Cracking (PWSCC)

PWSCC is not applicable as a potential damage mechanism for the RCS system due to the fact that there is no inconel(Ahoy 600) present in the system.

5.2.3 Localized Corrosion (LC)

In addition to SCC, other phenomena can produce localized degradation in piping components. These phenomena typicaHy 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 microbiologicaHy influenced corrosion, pitting, and crevice corrosion.

O Q 5.2.3.1 MicrobiologicaHy influenced Corrosion (MIC)

AH of the piping in this system, with the exception of the auxiliary spray line, operates in excess of the MIC upper temperature limit of 150*F. Consequently, the piping that operates at a temperatas above this limit is not considered susceptible to MIC attack.

The aaxiliary spray line operates at a temperature of 140*F. However, the VCT water supply, which serves as the source of water for auxiliary spray during plant cooldowns, is used for normal RCS makeup during charging operations. The excess system water, after having been exposed to RCS operating temperatures, is retumed to the VCT via letdown. Therefore, any microbes potentiaHy preseta on the VCT water supply would be eliminated by exposure to the elevated RCS operating temperatures. Consequently, the auxiliary spray line is also not considered susceptible to MIC attack.

5.2.3.2 Pitting (PIT)

The piping in this system is not exposed to oxygen or oxidizing species during normal power operation (RCS backpressure maintains isolation valves closed preventing oxygenated RWT water supply intrusion) and plant chemistry controls ensure that initiating contaminants (e.g., fluoride, chloride) levels are negligible.

Consequently, this system is not considered susceptible to pitting attack.

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%IFIF Calculation No. A PENG CALC 012, R2v. 00

' Page 30 of 49 5.2.3.3 Crevice Corrosion (CC)

Thermal sleeves, which are considered prime crevice region locations, are installed in the cold leg injection nozzles, hot leg surge nozzle, pressuritor surge nozzle, and the pressurizer spray nozzle. However, the piping in this system is not exposed to oxygen or oxidizing species during normal power operation (RCS backpressure maintains isolation valves closed preventing oxygenated RWT water supply intrusion). Consequently, this system is not considered susceptible to cnevice corrosion.

5.2.4 Flow Sensitive (FS)

When a high fluid velocity is combined with various other requisite factors it can result in the erosion and/or corrosion of a piping materialleading to a reduction in wall thickness. Mechanisms that are flow sensitive, and can create this form of degradation include crosion-ccvitation and flow accelerated cnosion.

5.2.4.1 Erosion-Cavitation (E-C)

All of the piping in this system, with the exception of the auxiliary spray line, operates in excess of the E C upper temperature limit of 250*. The auxiliary spray line however, which operates at a temperature of 140*F, experiences flow less than 100 hrs /yr, and the flow velocity is less than 30 ft/s. Furthermore, there are no potential sources of cavitation (e.g., pressure reducing orifices or valves) in the entire system. Consequently, this system is r.ot considered susceptible to E-C.

5.2.4.2 Flow Accelerated Corrosion (FAC)

The RCS system is comprised entirely of austenitic stainless steel piping, with the exception of the hot and cold legs which are carbon steel with a cladded austenitic stainless steelinternal surface (Reference 9.4). Since FAC is a phenomenon that only affects carbon steel piping, the RCS system is not susceptible to this ougradation mechanism (Reference 9.12).

5.2.5 Vibration Fatigue Vibration fatigue is not specifically made part of the EPRI risk-informed ISI process.

Most documented vibrational fatigue failures in power plants piping indicate that they are restricted to socket welds in small bore piping. Most of the vibrational fatigue damage occurs in the initiation phase and crack propagation proceeds at a rapid rate once a crack forms. As such, this mechanism does not lend itself to typicalperiodis inservice examinations (i.e., volumetric, surface, etc.) as a means of managing this degradation mechanism.

Management of vibrational fatigue should be performed under an entirely separate program taking guidance from the EPRI Fatigue Management Handbook (Reference 9.11). 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 .

ABB Combustion Engineering Nuclear Operations

A Ik k M HP O Calculation No. A PENG CAI.C-012. Rev. 00 Page 31 of 49 1 practical measures to address this issue. Because these measures are employed j either singly or in combination for most plant systems it is not necessary to use a 1

risk-informed inspection selection process for vibration fatigue.

5.3 BASIC DATA S.3.1 Under normal plant operating conditions, the RCS system, as defined by the boundaries in Section 3.2, functions as indicated in the table below.

Mping Une NormelMont Operating Number Description Temperetwe Flow Condtlons 2BCA 112' Pressurizer Surge line 611 Stagnant or low flow 2BC/.14 6' Pressurizer Vent Une (LTOP) 652 Stagnant or low Flow 2BCA 14-4' 2BCA 14 3' 2BCA-0-6' Pressurizar Safety Unes 652 Stagnant or low flow 2CCA 13 3* Pressurizer Main Spray Unes 553 Stagnant or Low Flow 2CCA.14 3' 2CCA 15 3' 2CCA 15-l*

\

( 2CCA 16-2* Pressurizer Auxiliary Spray Line 140 Stagnant or low Flow 2CCA.142' Hot legs 611 FullFlow 2CCA 2 42' 2CCA 3 30' Cold legs 553 Full Flow 2CCA-4 30' 2CCA 530' 2CCA-6 30' 2CCA 7-30' 2CCA 8 30' 2CCA 9-30' 2CCA.10-30

2CCA 29-2' Cold leg Drain Unes 553 Stagnant or low flow 2CCA.30-2' 2CCA-312

2CCA-32 2* Hot leg Drain Une 611 Stagnant or low Flow 5

5.3.2 Due to the cyclic nature of thermal transients, only those transients which occur during the initiating events Categorias I and ll as described in Reference 9.1, Table 3.1 are considered in the evaluation of degradation mechanisms due to thermal fatigue. Category I consists of those events which occur during routine operation, e.g., startup, shutdown, standby, refueling. Category 11 consists of those events w/olch have anticipated operational occurrence, e.g., reactor trip, turbine trip, loss of feedwater. Therefore, the transients to be evaluated are those transients which occur under normal operating and upset conditions.

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AhR M INIP f}

V Calculation No. A PENG-CALC 012, Rev. 00 Page 32 of 49 6.0 SERVICE HISTORY AND SUSCEPTIBillTY REVIEW An exhaustive review was conducted from mid '96 to Spring '97 of databases (plant and industry) and station documents to characterke 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 Reactor Coolant System.

Although several pre-commercial references are included for completeness, the time frame for identifying items applicable to this effort was focused on post-commercial operation (Commercial Operation date of March 26, 1980). This was done to avoidinclusion ofitems primarily associated with construction deficiencies as opposed to inservice degradation.

The following databases and other sources were queried to accomplish this review:

- Station Information Management System (SIMS)

The SIMS database was queried for all ANO-2 job orders on Code Class 1, 2, and 3 components which involved corrective maintenance (CM) or modifications (MQi)).

Additionally, s sepcrate query was performed in order to capture certain non-Code, O component failures. This query was for non-Code Q and SR (safety related) components. This database contains information from approximately 1985 to the present.

("

5.

- Condition Report (CR) Database The CR database was queried for any pipe leak / rupture events or other conditions associated with identified damage mechanisms at ANO 2. The keywords searched under were; pipe, piping, line, water hammer, leak, leaking and leakage. CR's are written on Q, F or S equipment failures or other conditions potentially adverse to safety.

This database contains information from 1988 to the present.

- Ucensing Research System (LRS)

The LRS database was queried using a keyword search specific to ANO-2. The keywords searched under were: thermal cycling, thermal stratification, thermal fatigue, defect, flaw, indication, fatigue, cavitation and corrosion. This search captured all communication between ANO and the NRC, both plant specific and gueric industry, associated with these topics. However, for the purpose of thii review, only communication from ANO to the NRC was reviewed. Additionally, th s search system was used to query Industry Events Analysis files (captures INPO doce.nents) for ANO-2 events or conditions relevant to this review. The keywords searched under for this portion of the query were: pipe & stratification, thermal & fatigue, thermal & transient, pipe & leak, vibration & fatigue and pipe & rupture. " Fuzzy" search logic was employed to reduce the possibility of failing to identify a pertinent document. This database contains information from prior to commercial operation to the present for ANO-2.

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ABB Combustion Engineering Nuclear Operations

ABB Calculation No. A.PENG. CALC 012, Rev. 00 Page 33 of 49

- Nuclear Plant Reliability Data System (NPROS)

NPROS was querled for ANO 2 entries for pipe failures. The keyword searched under t ras: pipe. This database contains Information from 1991 to the present.

- ANO 2 ISI Program Records The ISI program findings were conspiled and reviewed for all outage and non outage Inservice inspections conducted at ANO 2 since commercial operation.

- ControlRoom Station tog The station log was utilfred as a source of information for recent operational events.

The log exists in electronic format from early 1994 to the present and has search capabilities which allowed a review for events of interest. The keywords searched under were: water hammer, leak andleakage.

System Upper levelDocument (ULD)

The ULO was reviewed as a source for historical persoective of issues related to the system and identification of modifications made to .he sy3 tem or changes to operational procedures to address those issues (e.g., water hatamer, corrosicn or vibrational fatiguel.

~

9.0 REFERENCES

9.1

Risk. Informed Inservice Inspection Evaluation Procedure,' EPRI Report No. T".

106706, Interim Report, June 1996, 9.2 EPRIInservice Inspection Software llSIS*l,1996.

r 9.3 Arkansas Nuclear One Unit 2, " Safety Analysis Report,' Amendment No.13.

I 9.4

Design Specification for ASME Section til Nuclear Piping for Arkansas Nuclear One Unit :, Arkansas Power and Ught Company,' Specification No. 6600-M 2200, Revision 9.

9.5 *ANO 2 SIMS Components Database,'(Plant Pioing Une Ust (M 2083), dated 3 31-9 61.

9. 6 *ANO 2 ISI Plant Piping Une Ust,' from Res!sion 4 of ANO 2 Inservice Inspection Pla.-

9. 7 " Instruction Manual, Reactor Coolant Pipe and Fittings,' Arkansas Nuclear One Unit No. 2, C.E. Book No. 73470, June 1974.

9. 8

instruction Manual, Pressuriser,' Arkansas Nuclear One Unit No. 2, C.E. Book No.

p 73370, June 1974.

9.9

Technical Snecification for Insulation for Arkansas Nuclear One Unit 2 of the Arkansas Power and Ught Ccmnany." 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 VI, V2, V3, V4, Project 332101, Final Report, December 1994.

9.12

Pipe Cracking in PWRs with low Pressure Borated Water Systems,* EPRI Report No. NP-3320.

9.13 " Flow Accelerated Corrosion Prevention Program,' HES 05, Revision 1.

9.14 Arkansas Nuclear One, Unit 2, " Technical Specifications, Appendix A to Ucense No.

NPF 6, Amendments Nos.173 and 174.*

9.15 Gaertner, J. P., et. al., " Arkansas Nuclear One Unit 2 Internal Flood Screening Study,' prepsred 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.

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ABB Combustion Engineering Nuclear Operations

ABB

Calculation No. A PENG CALC 012, Rev. 00 Page 48 of 49 9.17 Entergy, Arkansas Nuclear One Unit 2, Drawings:

1. 0 Drawing No, M 2230 Sheet 1, Rev. 71, Sheet 2 Rev. 26;

Piping &

Instrumentation Diagram Reactor Coolant System.'

2.0 Drawing No. M 2231, Sheet 1, Rev.121; " Piping & Instrumentation Diagram Chemical & Volume ControlSystem.'

3.0 Drawing No. 2BCA 141, Sheet la Rev.15; "Large Pipe Isometric Reactor Coolant System.'

4. 0 Drawing No. C 6370-501 108, Rev. 01; ' Cold leg Zone 14 Steam l Generator #2 to Pump 28.* 1 5.0 Drawing No. C 6370 601 109: Rev. 01; ' Cold leg Zone 15 Pump 28 to Reactor Vessel.'

6.0 Drawing No. C ARK 511006;

Hot leg Zone 7 Steam Generator #2 to Reactor Vessel.'

7. 0 Drawing No. C ARK 511001; ' Cold leg Zone 8 Steam Generator #1 to Pump 1A. *

8. 0 Drawing No. C-ARK 511008; Rev. Oli ' Cold leg Zone 9 Pump 1A to Reactor Vessel.'

9.0 Drawing No. C ARK 511009; " Cold leg Zone 10 Steam Generator #1 to Pump 18. '

10.0 Drawing No. C ARK 511010; ' Cold leg Zone 11 Pump 18 to Reactor Vessel."

11.0 Drawing No. C ARK 511011; " Cold leg Zone 12 Steam Generator #2 to Pump 2A."

12.0 Drawing No. C ARK 511012; ' Cold leg Zone 13 Pump 2A to Reactor Vessel.'

13.0 Drawing No. D 6370 501001, Rev. 01; " Hot leg . Zone 6 Steam Generator

1 to Reactor Vessel."

14.0 Drawing No. D 6370 501023 Rev. 02: " Zone 16 Surge Une*, Une No.

2BCA 1 12.*

15.0 Drawing No. 2CCA 131, Sheet 1, Rev.13; "Large Pope isometric Reactor Coolant System."

16.0 Drawing No. 2CCA 141, Sheet 1, Rev.15; "Large Pipe isometric Pressurizer Spray System. "

17.0 Drawing No. 2CCA 151, Sheet 1, Rev. 9; *Large Pipe Isometric Discharge from Spray Valves to Pressurizer Spray System Header."

18.0 Drawing No. 2CCA 15 2, Sheet 1, Rev.14; 'Lorge Pipe Iscmetric Pressurizar Spray System."

19.0 Drawing No. 2CCA 15-4, Sheet 1, Rev. 4; "Large Pipe Isometric Pressurizer Spray System.

20.0 Drawing No. 2CCA 161 Sheet la Rev.17; *SmallPipe Isometric Chemical &

Volume ControlSystem.'

21.0 Drewing No. 2CCA 16 2 Sheet 1, Rev. 3; "Small Pipe isometric Auxiliary Spray from Regenerative Heat Exchanger. '

22.0 Drawing No. 2CCA 291, Sheet 1, Rev. 7; "Small Pipe isometric Reactor Coolant System Drain.'

23.0 Drawing No. 2CCA 30-1, Sheet 1, Rev. 7; "Small Pipe isometric Reactor Coolant System Drain to Reactor Coolant Tank 2T 68.'

24.0 Drawing No. 2CCA 31 1, Sheet 1, Rev. 6; "Small Pipe isometric Reactor Coolant System Drain to Reactor Drain Tank.'

ABB Combustion Engineering Nuclear Operations

ABB Calculation No. A PENG CALC 012, Rev. 00 (O3 Page 49 of 49 25.0 Drawing No. 2CCA 321, Sheet 1, Rev. 9; "Small Ppe isometric Reactor Coolant Drain System.'

9.18 *Pressuriser Surge Une Flow Stratification Evaluation,' CEOG Report No. CEN-387 P, Rev. IP-A, Afay 1994.

9.19 ' Temperature Distribution and Structural Analysis of Pressuriser Safety and Relief Valve Piping Sublect to Thermal Stratification,' CEOG Task 827, Report No. CE NPSD 1003, Revision 00, July 1995.

9.20

Thermal Stratification in Pressuriser AJxiliary Spray Piping,' CEOG Task 886, Report No. CE NPSD 1020, Revision 00, February 1996.

9.21

Stress Report for Automatic Pressuriser Vent Une & LTOP," Arkansas Nuclear One, Unit 2, Entergy Calculation No. 910 2016 04, Rev. 00.

9.22 'Pressuriser Spray System Thermalfatigue Evaluation,' CEOG Task 482, Report No.

CE NPSD 261, December 1984.

9.23 Swain, A. D. and Guttmann, H. E., 'Hondbook of Human Reliabidty Analysis with Emphasis on Nuclear Power Plant Operations,' NUREG CR 1278, August 1983.

9.24 Interoffice Correspondence from A. V. Bauer to Gustity Records, letter No. PENG.

O 97140, *Submittelof SIA Calculations,"datedJuly 21,1997.

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Calculation No. A MNG CALC 012, Mov. 00 Pnge A1 of A32 l l;.

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V APPENOlX A

FMECA CONSEQUENCEINFORMATION flCPOMT*

i (Attachment Pages /1. A32) ;-

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ABB Combustion Engineering Nuclear Operations

FMECA Consequence Infonnation Report Cah'a'"a Na A MGCAM 0". h 00 14-rar 91 Page A2 ef A.ft Consequence ID: RCS C 01 Consequence

Description:

loss of reactor coolant occurs via either RCS hot leg to the steam generators.

Break Stre: Large Isol.bility of Dreau No ISD Comments: The break is postulated to occur during normal power operation, and in the piping in either of the two RCS hot legs or the pressure surge line. This consequence evaluation includes all welds in lines 2CCA.142",2CCA , (2* & 2BCA l.12".

A failure in this segment would result in a large less of Coolant Accident (LOCA). Tius is characterir.ed by a rapid decrease in RCS pressure, followed by automade plant shutdown in order to bring the plant to a stable state. The lost RCS inventory would drain to the co.itainment surr.p as expected and would then be recirculated by the liPSI pumps. The failed segment cannot be isolated because there are no isoladon valves in the RCS hot legs.

Spatial Lifects: Containment Affected location: Containment Building Spatial Effects Comments: A dynamic analysis which included the above lines has been performed. The analysis concluded that in the case of a postulated guillotine break within the vessel subcompartment in either of the two hot leg pipe spools, the steam generator support system and the reactor vessel support system will limit the separation of the piping at the break and maintain the resulting subcompartment pressure transient within the design tiraits of the vessel support system (SAR Section 3.6.4.2.1.2).

Cenain safety related components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components include the contalmnent cooling units, SITS and their associated electrical equipment. The containment cooling units are designed to maintain their functional integnty following a breach of the RCS pressure boundary (SAR Section 6.2 2.2.2). In addition, all safety irdection components located inside the containment and their associated electrical equipment have been designed to withstand the LOCA emironment (S AR Section 6.3.2.12.1). llence, the impact of spatial effects due to the segment failure is assumed to be negligible.

Initiating Eventi l Initiating Event ID: A Initiating Event Recmcry: Based on the ANO 2 IPE (Report 94 R 2005 01, Rev. 0) two of four SITS, one of three HPSI pumps, and one of two LPSI pumps are required for successful mitigation of a large LOCA d.tring RCS inventory control (i.e., injection mode).

For long term inventory control and heat removal (i.e., recirculation mode), one of three HPSI pumps and one CS pump with an associated SDC heat exchanger or one CS pump and two containment cooling units are required for mitigating a l'uge LOCA. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiating event.

Loss of System: N System IPE ID: N/A System Recovery: During a large LOCA, the IIPSI, LPSI, SITS, CS, and Containment Cooling System will not be affected. No operator actions or automatir isolation are needed to recover from the segment failure. All engineered safety features required for mitigating this initiating event are actuated automatically.

4 O

p FMECA Consequme Infonnation Report Calmfah A'a A.PNOMJ, Rn M Q 1&sw91 Po p A 3 of A 32 Ims of Trale N TraleID: N/A Trale Recovery: N/A Consequence Consment: A large LOCA occurs due to a failure in the pipe segment. In Table 1 a LOCA is classified as a limiting fault event. Because the segmeni failure causes an initiating event, and based on Table I which was developed specift: ally for ANO 2 using the guidelines provided in Tables 3.1 mad 3.4 of the EPRI procedure (EPRI TR 106706), a HIGH consequence category is assigned.

Commequence Category: HIGH O co..e, ce Ra.k O O

O

FMECA Consequence Information Report Cah'aao"Na A FMG C4LC 0N R" 00 l4-sen Pop A4 of A.t2 l Consequence ID: RCS C 02 Consequence

Description:

Loss of reactor coolant occurs sia RCS cold leg 2P32 A.

l Break Size: Large isolability of Hrcak: No 150 Comments: The break is postulated to occur during normal power operation, and in RCS cold leg 2P32 A piping. This consequence evaluation includes all weldt in lines 2CCA 3 30" & 2CCA-4 30".

A failure in this segment would result in a large Loss of Coolant Accident (LOCA). T1ds is characterized by a rapid decrease in RCS pressure, followed by automatic plant shutdown in order to bring the plant to t stable state. The lost RCS imentory would drain to the containmet,t sump as expected and would then be recirculated by the lipSI pumps. The failed segment cannot be isolated because there are no isolatior valves in RCS cold leg 2P32A. RC5 makeup from the ECCS via RCS loop 2P32A will be ineffective.

Spatial Effects: Containment Affected location: Containment Building Spatial Effects Comments: A dynamic analysis which included the above lines has been performed. The analysis concluded tlut pipe motion will be limited because of restraints located in the pipe tunnels for e ch of the pump discharge pipes (SAR Section 3.6.4.2.1.2).

Certain safety related components uhich 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 LOC A (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). Ilence, the impact of spatial effects due to the segment failure is assumed to be negligible.

Initiating Esent: 1 Initiating Event ID: A Initiating Esent Reemcry: Based on the ANO 2 IPE (Report 94 R 2005 01, Rev. 3) two of four SITS irdecting into the intact RCS cold legs, one of three IIPSI pumps, and one of two LPSI pumps are required for successful mitigation of a large LOCA during RCS inventory cont ol (i.e., injection mode). For long term RCS imcr. tory control and heat removal (i.e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated SDC heat exchanger or one of two CS pumps and two containment cooling units are required for mitigating a large LOCA. Automatic actuation of the reactor protection system and the enginected safety features actuation system occurs in response to this initiating event.

Loss of System: SDM 3 System IPE ID: HPSI, LPSI, SIT System Recovery: During a large LOCA, the HPSI, LPSI, and SIT systems will operate in a degraded but effective manner. These systems will not fail to perform their intended design function (i.e.,

mitigation of a large LOCA) as a result of the segment failure. No operator actions or automatic isolation are needed in order to recover from the segment failure.

IAu of Train: N Train ID: N/A Train Recovery: N/A G

l

i FMECA Consequence Information Report Ca'c""* N" d"NGNdu R" 80 l4-sey91 Page A3 of A32 Co.neque.ee Comrigst: Following a large LOCA, makeuo from the ECCS via RCS loop 2P32 A is assumed to be ineffective due to a failure of the line ugment. ECCS flow to the remaining reactor l coolant loops will not be affected, and this level of performance is capable of mitigating a large LOCA. In Table 1, a LOCA is classified as a limiting fault emit 1 Because the segment failure causes an initiating event and degradation of Er.,

] frucction, and based on Table I which was developed specifically for ANO 2 w,ing tie i guidelines provided in Tables 3.1 aad 3.4 of the EPRI procedure (EPRI TR 106706), a 4

HIGH consequence category is assigned. ;

3 Co.neque.cc Category: HIGH O Co..eq.e.c. na.k O i

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. , - _ - . . . . , _ - _ - - __, . - , - . ~ . . - - . - ,, - . - . - _ - , - - _ . - . - -

FME'A - Consequence Infonnation Report Cak'da'='n A.Pruc cAte-on.sn oo l4-sey9) Pop A6 of AH Consequente ID: RCS-C 03 Consequence Description less of reactor coolant occurs via RCS cold leg 2P328.

Break Slee: Large Isolability of Break No ISO Comments: The break is postulated to occur during normal power operatica, and in RCS cold leg 2P32B piping. This consequence evaluation includes all welds in li.ies 2CCA 5 30" & 2CCA 6 30".

A failure in this segment would result in a large less of Coolant Accident (LOCA). Tids is characterized by a rapid cercase 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 pumps. The failed segment cannot be isolated because there are no isoladon valves in RCS cold leg 2P32B. RCS makeup from the ECCS via RCS loop 2P32B will be ineffective Spatial Effects: Containment Affected 14 cation: Containment Building Spatial Effects Comments: A dynamic analysis which included the above lines has been performed. The analysis concluded that pipe motion will be limited because of restraints located in the pipe tunnels for each of the pump discharge pipes (SAR Section 3.6.4.2.1.2).

Certain safety related components which are designed to mitigate the consequenec.

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), llence, the impact of spatial effects due to the segment failure is assumed to be negligible.

Initiating Esent: 1 Initiating Event ID: A initiating Event Reemcry: Based on the ANO 2 IPE (Report 94 R 2005-01, Rev. 0) two of four SITS injecting into the intact RCS cold legs, one of three IIPSI pumps, and one of two LPSI pumps are required for successful mitigation of a large LOCA during RCS inventory control (i c., injection mc.de). IW 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 associated SDC heat exchanger or one of two CS pumps ind two containment cooling units are required for mitigating a large LOCA. Automatic actuation of the reactor protection system and the engineered safety featA 's actuation system occurs in response to this initiating event.

Loss of System: SDM 3 System IPE ID: IIPSI, LPSI, SIT Sy stem Recmcry: During a large LOCA, the IIPSI, LPSI, and SIT systems will operate in a degraded but effective manner. These systems will not fail to perform their intended design function (i.e.,

mitiration of a large LOCA) as a result of the s.gment failure. No operator actions or 2

automat c isolation are needed in order to recover from the segment failure.

Loss of Train: N Train ID: N/A Train Recovery: N/A O

FMECA Consequence Information Report Ca'a'aa'*

A F M G 4 4LC 812

80 O 14-Ser>97 Pagt At of A32 Consequence Cosmewat Following a large LOCA, makeup from the ECCS via RCS loop 2P32B is assunal to be ineNective due to a failure of the line segment. ECCS flow to the remaining reactor coolant loops will not be aNected, and this level of performance is capable of mitigating a lange LOCA. In Table 1 a LOCA is classified as a limiting fault event.

Because the segment failure causes an initiating event and degradation of ECCS injection, and based on Table I which was developed specifically for ANO 2 using tiw guidelines provided in Tables 3.1 and 3.4 of the EPRI ;.rocedure (EPRI TR 106706), a lilGH consequence category is assigned.

Consequence Categor>1 HIGH O Conwguence Rank C O

O

FMECA Consequenee Infonnation Report Cabla'a xa A truc cac on, n,v. oo 14-sm91 Pog, A8 of A12 Consequence ID: RCS C-04 Consequence

Description:

Loss of reactor coolant occurs via RCS cold leg 2P32C.

Break Slie: Large isolability of Break No ISO Comments: The break is postulated to occur during normal power operation, and in RCS cold leg 2P32C piping. This consequence evaluadon includes all welds in lines 2CCA 7 30' & 2CCA 8 30",

A failure in this segment would result in a large Loss of Coolant Accident (LOCA), 'Ihis is characterized by a rapid decrease in RCS pressure, followed by automade plant shutdown in order to bring the plant tu a stable state. The lost RCS inventory would dram 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 vahts in RCS cold leg 2P32C. RCS makeup from the ECCS via RCS loop 2P32C will be ineffective.

Spatial Effects: Containment Affected Location: Containment Buildint Spatial Effects Comments: A dynamic analysis u hich included the above lines has been performed. The analysis concluded that pipe motion will be limited because of restraints located in the pipe tunnels for each of the pump discharge pipo (SAR Section 3.6.4.2.1.2).

Certain safety related components which are designed to midgate the consequences of a LOCA art also located inside the containment. T1.ese components include the containment coiling units, SITS a '. their electrical equipment. The containment cooling units are designed to m .in their func1onal integrity following a LOCA (SAR Section 6.L2.2.2). In addition, all safety irpection components located inside W

the containmem and their associated electrical equipment have been designed to withstand the LOCA emironment (S AR Section 6.3.2.12.1). llence, the impact of spatial effects due to the segment failure h assumed to be negligible.

Initiating Escot: 1 Initiating Event ID: A initiating E5ent Recovery: Based on the ANO-2 IPE (Report 94 R 2005-01, Rev. 0) two of four SITS injecting into the mtact RCS cold legs, one of three HPSI pumps, and one of two LPSI pump are required for successful mitigation of a large LOCA during RCS inventory control (i c., injection mode). I'or long term RCS inventory control and heat removal (i.e., recirculadon mode), one of three HPSI pumps and one of two CS pumps with an associated SDC heat exchanger or one of two CS pumps and two containment cooling units are required for mitigating a large LOCA. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiatir.g event.

Loss of System: SDM 3 System IPE ID: HPS1, LPSI, SIT Sy stem Recovery: During a large LOCA, the HPSI, LPSI, and SIT :ystems will operate in a degraded t,ut effective manner. These systems will not fail to perform their intended design function (i.e.,

mitigation of a large LOCA) as a result of the segment failure. No operator aedons or automatic isolation are needed in ordes to recover from the segmsnt failure.

less of Train: N Train ID: N/A Train Reemery: N/A O

N i

e FMECA Consequence Infonnation Report Ca'adaa*' n. A ruo cac on An. oo l<%e1 rage A9 of An Conwguence Comment: Following a large LOCA, nukeup from the ECCS sia RCS loop 2P32C is assumed to be ineffective due to a failure of the line ugment. ECCS flow to the remaining reactor a coolant loops will not be affected, and this level of perfornance is capable of j mitigating a large LOCA. i t Table 1, a LOCA is classified as a limiting fault event.

j Because the ugment fMlure causes an initiating event and degradation of ECCS j

injection, and based on Table I which t<as developed specifically for ANO 2 using the i

guidelines provided in Tables 3.1 and 3.4 of the EPRI procedure (EPRI TR.106706), a

HIGH consequence category is assigned.

Conwquence Cateson: HIGH O Co.wq.ence na.k O i

1 i

I l

I J

4 O

1 O .

FMECA Consequence Information Report Ca'ala'"a

d PENG CAM 812 ^ 00 l4-ser91 Page AIO of All Consequence ID: RCS C 05 Consequence

Description:

Loss of reactor coolant occurs via RCS cold leg 2P32D.

Break Size: Large Isolability of Break No ISO Comments: The break is postulated to occur during normal power operation, and in RCS cold leg 2P32D piping. This consequence evaluation includes all welds in lines 2CCA 9 30" & 2CCA.10 30",

A failure in this segment would result in a large Loss of Coolant Accident (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. The lost RCS inventory would drain to the containment samp as expected and would then be recirculatt ! by the IIPSI pumps. The failed segment cannot be isolated because there are no isolation vahts in RCS cold leg 2P32D. RCS makeup from the ECCS via RCS loop 2P32D will be ineffective.

Spatial Effects: Contahment Affected 14 cation: Containment Building Spatial Effects Comments: A oynamic analysis which included the above lines has been performed. The analysis concluded that pipe mu3on will be limited because of restraints located in the pipe tunnels for each of the pump discharge pipes (SAR Section 3.6.4.2.1.2).

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 cooUng units, SITS and their ciectrical equipment. The containment cooling units are designed to maintain their functional integrity following a LOCA (SAR Section 6.12.2.2). In addition, all safety i: Vection components located inside the containtnent and their associated electrical equipment hast been designed to withstand the LOCA emitonment (SAR Section 6.3.2.12.1). Ilence, the impact of spatial effects due to the segment failure is assumed to be nagligible.

Initiating Esent: I Initiating Event ID: A initiating Esent Recovery: Based on the ANO 2 IPE (Report 94 R 2005 01, Rev. 0) two of four SITi injecting into the intact RCS cold legs, one of three llPSI pumps, and one of two i. PSI pumps are required for successful mitigation of a large LOCA during RCS inventory control (i c., ir0cction mode). 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 SDC heat exchanger or one of two CS pumps and two containment cooling units are required for mitigating a large LOCA. Automatic actuation of the reactor protection system and the enginected safety features actuation system occurs in response to this initiating event.

less of Splem: SDM 3 Sptem IPE ID: IIPSI, LPSI, SIT Sy stem Reemery: During a large LOCA, the HPSI, LPSI, and SIT systems will operate in a degraded but effective manner. These systems will not fail to perform their imended design function (i.e.,

mitigation of a large LOCA) as a result of the segment failure. No operator actions or automatic isolation are needed in order to recover from the segment failure.

Loss of Train: N Train ID: N/A Train Recovery: N/A O

r FMECA - Consequence Information Report Calcularb:do. A PEAU 44LC-012,Rev 00

\ 14 sw91 Page All of A32 Consequence Comrru Following a large LOCA, makeup from the ECCS via RCS loop 2P32D is assumed to be ineffective due to a failure of the line regment. ECCS flow to the remaining reactor coolant loops will not be affected, and this level of p:rformance is capable of mitigating a large LOCA. In Table 1, a LOCA is classified as a limiting fault event.

Because the segment failure causes an initiating event and degradation of ECCS itsection, 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 TR 106706), a HIGH consequence category is assigned.

Consequence Category: HIGH O Consegmence aank O O

6

l FMECA - Consequence Infomtation Report Calculaa n Na ,f PENG-CA.C 012. Rev 00 14-sey91 Page A12 of A32 Consequence ID: RCS-C 06 Consequence

Description:

Loss of reactor coolant occurs via pressurizer maia spray line.

Break Size: Large Isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation,and in the main spray line piping - from the RCS cold legs (2P32A & 2P37B) to the pressurizer spray nozzle. This consequence evaluation includes all welds in lines 2CCA 13 3",2CCA 14 3",2CCA 15 3",

and 2CCA 15-4",

A failure in this segment would result in a medium Loss of Coolant Accident (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. The lost RCS im entory would drain to the containment sump as expected and would then be recirculated by the HPSI pumps. No isolation of the failed segment is assumed because there are no isolation valves for portions of the main pressurizer spraylines.

Spatial Effects: Containment Affected Location: Containment Building Spatial Effects Comments: A dynamic analysis (S AR Section 3.6.4.2.11.2) which included the above lines has been performed. The analysis concluded that the cavity walls (i.e., the steam generator cavity) containing the main pressurizer spray lines provide protection for the containment liner and the required systems outside the cavity, inside the cavity, the major equipment are protected from jet impingement. Restraints are also in place to guard against pipe whip.

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, SIT, and their associated electrical equipment. The containment cooling units are designed to maintain their ftmetional integrity following a breach in the RCS pressure boundary (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 a segment failure is assumed to be negligible.

Initiating Esent: I Initiating Event ID: M Initiating Event Recovery: Based on the ANO-2 IPE (Report 94 R 2005-01, Rey,0)one of three HPSI pumps is required for successful mitigation of a medium LOCA during RCS inventory control (i.e., injection mode). 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 cooling units are required for mitigating a medium LOCA. Automatic actuation of the reactor protection system and the enginected safety features actuation system occtrs in response to this initiating event.

Loss of S) stem: S System IPE ID: Main Spray System Recovery: No recovery of the main spray system is needed because this system is not required to mitigate the eficcts of a medium LOCA. All enginected safety features required for mitigating this initiating event are automatically actuated.

.. ._. _.____ _ _ _ _ _ . _ _ _ _. _ ._ _ _ . _ ___. _ _ . _ . ~ ._ ____ -._ __

p FMECA - Consequence Information Repot t l4-spe7 Caltsleon A'a A PEA'GN4/2.Rn. 00 Page A13 of A32 Loss of Train: N Train ID: N/A Traia Recovery: N/A Consequence Comment: The portion of this segment between the motor operated valves (MOVs) in the main spray lines can be isolated by closing the associated MOVs. A failure in other portions of this segment would result in rapid depressurization of the RCS which is indicated by

, the pressurizer pressure instruments. The indications and alarms in the control room would not provide the type of detailed information needed by the operators to determine that the break exists between the MOVs. Since the exact break location cannot be easily determined, no credit is taken for isolating a segment failure between the MOVs. A failure in any portion of this consequence segment is therefore treated as a medium Loss of Coolant Accident (LOCA) because of the pipe size.

Because the segment failure causes an initiating event, and based on Table I which was developed specifically for ANO-2 using the guidelines prosided in Tables 3.1 and 3.4 of the EPRI procedure (EPRI TR.106706), a HIGH consequence category is assigned.

Consequence Category: HIGH C Consequence Rank O i

i e

O 1 V 1

..-y

I l

FMECA - Consequence Information Report C* datum & A-PEAN 012. Rn 00 <

14-Sero i Page A14 of A32 i Consequence ID: RCS C-07 Consequence

Description:

Loss of reactor coolant occurs via pressurizer auxiliary spray line.

Break 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 the auxiliary spray check vahe 2CVC-28A to the main spray line. This consequence evaluation includes all welds in lines 2CCA 16 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 autom. tic plant shutdown in order to bring the plant to a stable state. The lost RCS imentory would drain to the containment sump as expected and would then be recirce'ated by the HPSI pumps. The failed segment cannot be isolated because there are no isolation s ahts in the abost line segment.

Spatial Effects: Containment Affected Location: Containment Building Spatial Effects Comments: A dynamic analysis (SAR Section 3.6.4.2.9.2) which included the abost 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 whip of this line. Restraints in addition to existing piping and structures will protect the shutdoun cooling vahrs and ensure that shutdown cooling is available if requirei 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 Event: 1 Initiating Event ID: S Initiating Event Recovery: Based on the ANO-2 IPE (Report 94-R 2005-01, Rev. 0), the follening 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 inventeuntrol (i.e., injection mode), one of three HPSI pumps.

(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 shtudown cooling heat exchanger or one of two CS pumps and two containment cooling units.

If a total loss of feedwater occurs, core heat terroval can be accomplished by "once through cooling" using the LTOP or ECCS vent valves. Automauc actuation of the reactor protection system and the engin:ered safety features actuation system occurs

FMECA - Consequence Information Repon Calculatum Na A PENG CGC 0M. Rn. 00

, 14 sg91 Pogs A13 of A32 in response to this initiating ewnt.

Less of S) stem: S Systein IPE ID: Aux Spray System Recovery: No recovery of the auxiliary spray system is needed because this system is not required for mitigating a small LOCA. All engineered safety features required for mitigating this initiating event are automatically actuated.

Iass of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: A small Loss of Coolant Accident (LOCA) occurs due to a failure in the line segment, in Table 1, a LOCA is classified as a limiting fault event. Because the segment failure causes an initiating 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 TR 106706), a HIGH consequence category is assigned.

Consequence Category: HIGH O Consequence aank O O

O

FMECA - Consequence Information Report Calculanon No A-PENG CALC 012. Rrv 00 14-sep.91 Page A16 of A.12 Consequence ID: RCS-C-08 Consequence

Description:

Loss of reactor coolant occurs via pressurizer pressure rehef/LTOP line.

Break Size: Large isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation, and in the piping from the pressurizer to upstream of LTOP valves 2CV 4730-1 & 2CV-4740-2, and to upstream of the PSVs. This consequence evaluation includes all welds in lines 2BCA 14-4",2BCA 14 6" and 2BCA 0-6".

A failure in this segment would result in a large Loss of Coolant Accident (LOCA). This is characterized by a rapid decrease in RCS pressure, followed by automatic plant shutdow11 in order to bring the plant to a stable state. The lost RCS inventory would drain to the containment sump as expected and would then be recirculated by the HPSI pumps. The failed segment cannot be isolated because there are no isolation vahrs in this portion of the pressurizer pressure relief /LTOP line.

Spatial Effects: Containment Affected Location: Containment Building Spatial Effects Comments: A dynamic analysis (S AR Sec; ion 3.6.4.2.12.2) which included the above lines has been performed. The analysis concluded that a failure in the line segment will not result in a more sestre LOCA than the design basis event it initiates.

Certain safety related components uhich are designed to mitigate the consequences of a LOCA are also located inside the containment. These componena include the containment cooling units, SITS and their electrical equipment. The containment cooling units are designed to maintain their functional in*egrity 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 Event: 1 Initiating Event ID: A Initiating Event Recovery: Based on the ANO-2 IPE (Report 94-R 2005-01, Rev. 0) two of four SITS injecting into the intau RCS cold legs, one of three HPSI pumps, and one of two LPSI pumps are required for successful mitigation of a large LOCA during RCS imentory control (i.e., injection mode). For long term RCS inventory control and heat removel (i.e.. recirculation mode), one of three HPSI pumps and one of two CS pumps with at, associated SDC heat exchanger tw one of two CS pumps and two containment cou.. g units are required for mitigating a large LOCA. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiating esent.

Loss of System: SM-3 System IPE ID: Aux Spray, Main Spray, PZR Vent System Recovery: No recovery of the pressurizer pressure control systems (i.e., Aux Spray, Main Spray, &

Pressurizer Vent) is needed because these systems are not required to mitigate the effects of a large LOCA. All engineered safety features required for mitigating this initiating estnt are automatically actuated.

O l

Calafatim Na A PENG<.4LC-0U, Rn. 00 FMECA - Consequence Information Report O l4-ser 91 Loss of Train: N Train ID:

Page Ali of An Train Recoscry: N/A Consequence Comment: A large LOCA occurs due to a failure in the line segment. In Table 1, a LOCA is classified as a limiting fault event. Because the segment failure causes an initiating 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 TR 106706), a HIGH consequence category is assigned.

Consequence Category: HIGH C Consequence Rank O O

FMECA - Consequence Information Report Catalam h A.PNGCMA N t 4-ser-97 Page AIB of A32 Consequence ID: RCS-C 09 Consequence

Description:

Loss of reactor coolant occurs sia RCS hot or cold leg drain line.

Break Size: Large isolability of Break: No ISO Comments: The break is postulated to occur during normal power operation, ar. a the piping from upstream of RCS drain valves 2RC 4B,2RC-4C,2RC-4D, or 2RC-4E to the RCS hot / cold I y This consequence evaluation includes the welds in the applicable portions oflines 2CCA 29 *,

2CCA 30-2",2CCA 312", & 2CCA 32 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 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 in the line segment.

Spatial Effects: Containment A.'lected Location: Containment Building Spatial Effects Comments: A dynamic analysis (SAR Section 3.6.4.2.1.2) has been performed for a break in the RCS cold leg. The analysis concluded that systems required to mitigate the consequence of the break will not be impaired byjet impingement or uncontrolled whip of this line. Since the dynamic effects of a cold leg break bound the effects of a break in the cold leg drain line, it is assumed that jet impingement and pipe whip will not be a concern due to a failure in the above segment.

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 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 c., injection mode), one of three HPSI pumps.

(c) For long term RCS inventory control and heat removal (i.e., recirculation mode), one of three HPSI pump 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 total 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

FMECA . Consequence Information Report Ca'cala'io No. A PENG<.ALC-012 Rev. 00

.O - :4-Ser 91 Page Al9 ofA32 i

reactor protection system and the enginected safety features actuation system occurs in response to this initiating event.

Less of System: N System IPE ID: N/A System Recovery: All enginected safety features required for mitigating this initiating ennt are automatically actuated. No operator actions are needed in order to recover from the segment failure.

Loss of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: A small Loss of Coolant Accident (LOCA) occurs due to a failure in the line segment.

in Table 1, a LOCA is classified as a limiting fault event. Because the segment failure d

causes an initiating 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 TR-106706), a HIGH consequence category is assigned.

Consequence Category: HIGH O Consequence Ra k O i

f 4

e i

4

FMECA - Consequence Information Report Caka'ation No. A PEAU-CALC-012, Rev. 00 14.ser 91 Page A20 of A32 Consequence ID: RCS-C 10 Consequence

Description:

Potential loss of reactor coolant occurs sia hot leg drain line.

Break Size: Large Isolability of Break: No ISO Comments: The segment failure is postulated to occur in the piping between hot leg drain line manual valves 2RC-4E and 2RC 5E. The piping is isolated during normal power operation. If drain valve 2RC-4E 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 4.1.7) 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 32 2".

Failure of drain valve 2RC-4E 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. The lost 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 vahts upstream of valve 2RC-4 E.

Spatial Effects: Containment Affected 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 (SAR Section 6.2.2.2.2). In addition, all safety injection components located inside the containment and their associated electrical equipment hast 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 Evant: 1 Initiating Event ID: S Initiating Esett 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.e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shtudown cooling heat exchanger or one of two CS pumps and two containment cooling units.

Calculan n No. A PENG CALC-012, Rrv. 00 FMECA - Consequence Infonnation Report O 14-sep 9, Page A21 of A32 If a total 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.

less of System: N System IPE ID: N/A Systen Recoveryt N/A less of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: Failure of drain valve 2RC-4E 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 drain valve and the conditional core damage probability for a small LOCA (Table 1) results in a MEDIUM consequence (see Section 4.1.7).

Consequence Category: MEDIUM O consequence Rank O O

O .

FMECA - Consequence Information Repon c.ai.n A'a ammc-o12 Rn. 00 14-sep.97 Page A22 of A32 Consco sence ID: RCS-C Il Consequence

Description:

Potential loss of reactor coolant occurs via cold leg 2P32B drain line.

Break Size: Large Isolability of Break: No ISO Comruents: The segment failure is postulated to occur in the piping bety an cold leg drain line manual valves 2RC-4B and 2RC 5B. The piping is isolated durint wrmal pcwcr operation. If drain valve 2RC-4B remains closed, a segment failure would not cause an initiating event. The drain line is r.lso 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 tsee Section 4.1,7) 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 29 2",

Failure of drain valve 2RC-4B followed by a segment failure would result in a small LOCA.

This is characterized by a rapid tiecrease 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 HPSI pumps. The failed segment cannot be isolated because there are no isolation vahu upstream of valve 2RC-4 B.

Spatial Effects: Containment Affected 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 irppaired byjet impingement or uncontrolled whip of this drain line. a Certain safety rela".d components which are designed to mitigate the consequences of a LOCA are also located inside the containment. These components inlude t..e containment cooling units, SITS, and their electrical equipment. The containment cooling units are designed to maintain thea 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 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 F.FW pumps or one of tvu main feedwater pumps.

(b) For RCS inventory control (i.e., injection mode), one of three HPSI pumps.

(c) For long term RCS inv ntory co . trol and heat removal (i.e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shtudomi cooling heat exchan:;cr or onc of two CS pumps and two containment cooling units.

FMECA - Consequence Infonnation Report Ca'a'aa* Na A SN-812 Ra #

!4-sep 91 Page A23 of A32 If a total 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 enginected safety features actuation system occurs in response to this initiating event.

Ius of System: N System IPE ID: N/A Synem Recoveryt N/A Lois of Train: N Train ID: N/A Tralu Recovery: N/A Consequence Comment: Failure of drain valve 2RC-4B to remain closed uill 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 drain valve and the conditional core damage probability for a small LOCA (Tabic 1) results in a MEDIUM consequence (see Section 4.1.7).

Consequence Category: MEDIUM O Consequence Rank O i

i O

i l

FMECA - Consequence Information Report Calmla w h A P N C E C-012 3 m 00 14-ser-91 Page A24 of A32.

Consequence ID: RCS C 12 Consequence

Description:

Potential loss of reactor coolant occurs via cold leg 2P32C drain line.

Break Size: Large Isolability of Break: No ISO Comments: The segment failure is postulated to occur in the piping between cold leg drain line manual valves 2RC-4C and 2RC-5C. The piping is isolated during normal power operation If drain valve 2RC-4C remains closed, a segment failure would not cause an initiating event. The dnin 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 4.1.7) 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 30-2".

Failure of drain valve 2RC-4C 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. The lost 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 vah es upstream of valve 2RC-4C.

Spatial Effects: Containmcnt Affected Location: Containment Building Spatial Effects 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 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 contamment cooling units are designed to maintain tlicir 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 Event: I initiating Even: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:

(0) For RCS and core heat removal, one of two EFW pumps or one of two main feedwater pumps.

(b) For RCS inventory ccattol (i.e., injection mode), one of three HPSI pumps.

(c) For long term RCS im entory control and heat removal (i.e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shtudown cooling heat exchangcr or one of two CS pumps and two containment cooling units.

[

( FMECA - Consequence Information Report 14-Sey91 Calculanon No. A-PENG-CALC 012 Rev 00 Pop A23 of A32 If a total loss of feedwater occurs, core heat removal can be accomplished by "once through cooling" using the LTOP or ECCS vent vahes Automatic actuation of the reactor protection system and the engineered safeiy features actuation system occurs in response to this initiating event.

Loss of System: N System IPE ID: N/A Syuem Recovery: N/A Loss of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: Failure of drain valve 2RC-4C 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 drain valve and the conditional core damage probability for a small LOCA (Table 1) results in a MEDIUM consequence (sm Section 4.1,7).

Consequence Category: hEDIUM O Consequence Rank O O

o v

FMECA - Consequence Infornstion Report Cahlation Na A PENG. CALC-0M.Rev 00 14-sep.91 Page A26 of A32 Consequence ID: RCS-C-13 Consequence

Description:

Potential loss of reador coolant occurs via cold leg 2P32D drain line.

Break Size: Large Isolability of Break: No ISO Comments: The segment failure is postulated to occur in the piping between cold leg drain line manual valves 2RC-4D and 2RC-5D. The piping is isolated during normal power operation. If drain valve 2RC-4D remains closed, a segment failure would not cause an initiating esent. The drain line is also not needed to support or accomplish any of the safety functions required for mitigating a design basis estnt. The resulting consequence would be less sigrlficant than the consequence of a small LOCA (see Section 4.1.7) 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-31-2",

Failure of drain valve 2RC-4D 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. The lost RCS imtatory 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 vahrs upstream of vahr 2RC-4D.

Spatial Effects: Containment Affected 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 impmgement 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 maletain their functional integrity foliowing a LOCA (SAR Section 6.2.2.2.2). In addition, all safety injection components located inside the containment and their associated electri:.al equipment hast 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 imentory control and heat removal (i.e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shtudown cooling heat exchanger or one of two CS pumps and two containment cooling units.

FMECA - Consequence litformation Report Catalan n No. A-FENG-CALC 012.Rev,00 14-$sp91 Pop A27 of A32 If a total loss of feedwater occurs, core heat removal can be accomplished by "once Caough 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.

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: Failure of drain valve 2RC-4D to remain closed will expose the segment to the operating tem.wratures 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 drain vahc od the conditional core damage probability for a small LOCA (Table 1) results in t. E- 3 IUM consequence (see Section 4.1,7).

! Consequence Category: MEDIUM O Cocaw 3ee aank O 1

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FMECA - Consequence Information Report Cabidon %. A PENG C4LC 012, Rev. 00 gs ser 91 Page A28 of A32 Conscauence ID: RCS-C 14 Consequence

Description:

Loss of auxiliary epray capability occurs due to a line bra.ak.

Break Size: Large Isolability of Break: No ISO Comments: The break is postulated to occur in the auxiliary spray line, upstream of check valve 2CVC-28A and downstream ofisolation valve 2CV-4824-2. The segment is normally isolated during power operation. If the check valve remains closed, a segment failure would not cause an initiating event. However, the auxiliary spray system would be lost when demanded. Because of detection capabilities and other available backup trains (i.e., LTOP and ECCS vent paths) for reducing RCS pressure, the resulting ennsequence is less significant than a small LOCA (see Section 4.1.7) 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 16-2".

Failure of check val <c 2CVC 28 A followed by a segment fai'ure 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. The lost RCS inventory would drain to the containment sump as expected and would then be recircula'ed by the HPSI pumps. The failed segment cannot be isolated because there are no isolation valves upstream of check valve 2CVC 28A.

Spatial Effects: Containment Affected Location: Containment Buildir.g 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 consequences of the break will not be impaired byjet impingement or uncontrolled whip of the line, it is therefore assumed that the dynamic effects of the segmer.t failure would nct tx a concern.

Initiating Event: 1 Initiating Event ID: S Initiating Event Recovery: Based on the ANO-2 IPE (Report 94-R 200541, 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 cor. trol and heat removal (i.e., recirculation mode), one of three HPSI pumps and one of two CS pumps with an associated shtudown cooling heat exchanger or one of two CS pumps and two containment cooling units.

If a total 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.

Iess of System: S System IPE ID: Aux Spray O

-. . - . _ - - _ - . _ _ . - . . . - _ = _ = - - . . . - - - ._._ -.-. -

J Catalanon Na AM612, Rn. M

O FMECA - Consequence Information Report 14-sep 91 System Recovery: No recovery of the auxiliary spray system is expected. Although the auxiliary spray sys nY Page A29 of A32 unavailable, RCS pressure can be reduced by using the ECCS or LTOP vent vahrs.

. Loss of Train: N Train ID: N/A Train Recovery: N/A +

l Consequence Comment: Failure of check valve 2CVC 28A 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 drain valve and the conditional core damage probability 4

for a small LOCA (Table 1) results in a MEDIUM consequence (see Section 4.1,7).

Consequence Category: MEDIU!, C Consequence Rank C O

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FMECA - Consequence Information Report Cahlata No. A PENG-CALC-012, Rev. 00 14-sep 91 Page A30 of A32 Consequence ID: RCS-C 15 Consequence Descrip' ion: Potential loss of reactor coolant occurs via LTOP/ECCS Vent Line.

Break Size: Large Isolability of Break: No ISO Comments: The break is postulated to occur in the piping from downstream of LTOP vahts 2CV-4730-1 and 2CV 4740 2 to upstream of LTOP valves 2CV-47312 and 2CV-4741-1. The ECCS vent line upstream of vent valve 2CV-4698 1 is also included in this line segment. The piping is 1

isolated during normal power operation. If motor operated valves 2CV 47301 and 2CV-4740-2 remain closed, a segment failure would not cause an initiating event. The ability to perform

" Feed and Bleed" would also not be impacted when demanded. The resulting consequence would be less significant than the consequence of a medium LOCA (see Section 4.1.7) wh:ch challenges the mitigating system. Therefore, only the case for a medium LOCA is described herein. This consequence evaluation includes the welds in the applicable portions oflines 2BCA 14-3" and 2BCA 14 4" Failure of motor operated valve 2CV-4730-1 or 2CV f 740-2 followed by a segment failure would result in a medium LOCA. This is characterized by a . apid 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 HPSI pumps following makeup from the Sfis and RWT, The failed segment cannot be isolated because there are no isolation valves upstream of moter operated vahes 2CV-47301 and 2CV-4740-2.

Spatial Effects: Containment Affected Imcation: Containment Building Spatial Effects Comments: A dynamit .aalysis (S AR Section 2.6.4.2.12.2) which included the above lines has been performed. The analysis concluded that a line break would not result in an event more severe than the design basis loss of coolant accident. Depending on the break location, the analysis showed that no adjacent safety related equipment would be affected or else an alternate path to safe shtitdown was identified in the event that potentially affected safety related equipment would become unavailable. It is therefore assumed that the dynamic effects of the segment failure y ould be negligible.

Initiating Event: 1 Initiating Event ID M Initiating Event Recos cry: Based on the ANO 2 IPE (Report 94-R 2005-01, Rev. 0) one of three HPSI pmnps is required for successful mitigation of a medium LOCA during RCS inventory control (i.e., injection mode). For long term RCS inventory control and heat removal (i.e.,

recirculation mode), one of tirtee 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 are required for mitigating a medium LOCA. Automatic actuation of the reactor protection system and the engineered safety features actuation system occurs in response to this initiating event.

. IAss of System: N . System IPE ID: N/A System Recovery: N/A less of Train: N Train ID: N/A O

J FMECA - Consequence Information Report Ca'a'a'""' A*o. mAUN011 Rn @

/

14.S997 Page A't of U.'

Train Recovery: N/A Consequence Comment: This segment is normally isolated during power operation. A failure of the interfacing LTOPECCS valve to remain closed will therefore expose the segment to the operating

< temperatures and pressures of the RCS. By exposing the segment to normal operating i conditions, the potential for a medium LOCA exists. The combined effect of a passive failure of the LTOPECCS valve and the conditional core damage probability of for a medium LOCA (Table 1) results in a MEDIUM consequence (see Section 4.1.7). ,

Consequtare Category: MEDIUM C Consequence Rank O j .

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FMECA - Consequence infecreike Report rahdavr& ANC4 Loo 12 Av.00 14Sep qi Awe A32 of A32 Tel ASSIGNED CONSEQUENCE CA1TGORIES FOR ANO2INmATINGEVENTS i.nau =i lanu.u.e Ev. lanauner ttu.capu rE cDr ccDr c-(1y) (CDF /IE Fr.q )

1 Rouuns Starse N'A N/A N 'A N/A Sksemn 8tandby Refuehng

!! Anucipaa.d Reaciar Tnp .(T6) 2 03 3 95E46 2 93E46 MEDIUM Lens of Power Comenai Sysiam. (T2) 0 25 I 99E-07 339E46 MEDIUM Twkne inp.(TI) 0 76 177E46 2 98E46 MEDIUM

!!! Infrequena Loss of Othne Po.w . (T3) $ 84E42 172E46 2 95E45 MEDIUM Less of SW Pwnp 2P4A .(T8) 738E42 214E-07 290546 MEDIUMM Loso of SW Pwnp 2P48.(T9) 738E42 2 04E47 2 77E46 MEDIUM "

IV Latuung Fauhs Excessrve Fe.dwater . (T4) 9 40E44 I t7E 09 199E4 MEDIUM or Accdents Stearn Lum 7eedwater Luis Br.ak . (TS) I 10E-03 113E49 I 03E46 MEDIUM Total Less of SW-(T7) $ 45E 03 214E 4 3 93EM 100H Less of DC Bus 2D01. (TIO) 3 94E44 980E 4 2 49E42 FOOH#

Loss of DC Bus 7D02. (Tl1) 394E M 1.11E-06 2 83E 03 1DOH

Less of AC Bus 2A3.(T12) 3 94Ea 3 23E-06 8 20E43 IDOH8' Loss of AC Be 2A4. (Tl3) 3 94E44 5 78E-08 I47E M FOOH" Loss of 480V LanJ Center 185 (T14) 104E-03 I 90E-01 183E44 10GH

Less of 480V Lo.J Center 2B6.(T15) 104E43 120E47 145E44 10GH*

Small LOCA.(S) 5 00E43 171E46 3 43Em 100H Mediurn LOCA -(M) 100E43 174E 4 174E43 }0GH Large LOCA -( A) 100E M 139E46 139E42 >0011 Steam 0cnmier Tube Rupewe . (R) 9 77E43 9 54E48 9 76E-06 kE.DIUM NOTE:

1. Thu into.u.s m r.s.sts in n mes.,irip d i .t iria .t r. ice woen,i. r.sr i ds.

t 2. This intu.u.g m.t re hs in a r ct.c irty d peru.! i.es t effsit. pow r.

3. Th. i.au.u.s m. r h. i. . met.,irip ..d in .t p.wm. w.6. .f unius.u.s sy.i

IIIGII" CCDP>10 d d

" MEDIUM" 10 < CCDP < 10d

" LOW" CCDP < It' The above table was developed for the ANO-2 specific initiating events. It is based on the informaation provided in Tables 3.1 and 3.4 of &

EPRI RISI procedure (EPRI TR-106706). The initiating event descriptions (for event categories II, III, & IV) and associated in6 stating event frequencles and Core Densage Frequencies (CDFs) wen extracted fresa Tables 3.3-6 and 3.5.4-7A of the ANO-2 IPE (Report 94-R.

2005-01, Rev. Oh O

1 1.

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! Calculation No. A PENG-CALC 012, Rev 00 Page 81 of 836

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! APPENDIX 8 1

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t "FMECA - DEGRADA TION MECHANISMS' t

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1 (Attachment Pages B1 - 836) 1 i

I t,

ABB Combustion Engineering Nuclear Operations

'" FMECA - Degradation Mechanisms cara, tan n Aa Ammai2.Ru. oo Page B2 of B36 4 Weld l 3ptem ID Segment Line Nunnber Line Description Number Weld Iecation T C P I M E F O RCS RCS-001 2BCA-0-6" Pressurizer Safety 42-002 Weld at 2PSV-4633 nozzle No No No No No No No No Vahr inlet line RCS RCS-001 2BCA-0-6" Pressurizer Safety 42-003 Weld at 2PSV-4634 nozzle No No No No No No No No Valve inlet line RCS RCS-001 2BCA-14-6" Pressurizer vent line 43-021 Downstream of 6* tee #5 No No No No No No No No (flange side)

RCS RCS-001 2CCA-1-42" Ilot leg from reactor 06-001 Circumferential utid in No No No No No No No No vessel to steam Ieop A of RCS hot leg generator 2E24A downstream from reactor vessel RCS P.CS-001 2CCA-t-42" Ilot leg from reactor 06-002 Circumferential weld in No No No No No No No No vessel to steam Loop A of RCS hot leg generator 2E24A dowr. stream from reactor wssel RCS RCS-001 2CCA-1-42" liot leg from reactor 06-005 Circumfe.ential weld in No No No No No No No No vessel to steam Imp A of RCS hotleg generator 2E24A downsticam of pressurizer surge line RCS RCS-001 2CCA-I-42" Ilot leg from reactor 06-008 Circumferential weld in No No No No No No No No vessel to steam loop A of RCS hotleg generator 2E24A dowmstream from pressurizer surge line RCS RCS-001 2CCA-I-42" flot leg from reactor 06-009 Circumferential utld in No No No No No No No No vessel to steam loop A of RCS hot leg generator 2E24A next to steam generator 2E24A Dearadation Mechanisms T-Thermal Fatigue P - Ihmary Water stress Corrosion Cracking (PWSCC) M - Microluologicatty inHuenced 8 -mian(MIC) F- Flow treelerused Cornsoson C-Corrosson Crackmg I- hnergranular Stress Comnion Cracking CGSCC) E - Erasue-Caritason 0-Other O O O

f n - m -

"**P'7 FMECA - Degradation Mechanisms C""#""

" #" "",, 2;R Weld System ID Sepnent Line Number Line Description Number Weld Location T C P I M E F 0 RCS RCS-001 2CCA-I-42" Ilot leg from reactor 06-010 Branch connection weld in No No No No No No No No vessel to steam Loop A of RCS hot leg for generator 2E24 A pressurizer surge line RCS RCS4)01 2CCA-1-42" Hot leg from reactor 06-011 Branch connection stid No No No No No No No No vessel to steam for Loop A of RCS hotleg generator 2E24A drain line RCS RCS-001 2CCA-10-30" Cold leg from reactor 15-001 Circumferential weld in No No No No No No No No coolant pump 2P32D to RCS cold leg next to reactor vessel reactor sessel(discharge side of RCP 2P32D)

RCS RCS-001 2CCA-10-30" Cold leg from reactor 154)02 Circumferential weld in No No No No No No No No coolant pump 2P32D to RCS cold leg (discharge reactor vessel side of RCP 2P32D)

RCS RCS-001 2CCA-10-30" Cold leg from reactor 15-005 Circumferential weld in No No No No No No No r4o coolant pump 2P32D to RCS cold leg downstream reactor vessel of Siline (discharge side of RCP 2P32D)

RCS RCS-001 .2CCA-10-30" Cold leg from reactor 154)c8 Circumferentia: weld in No No No No No No No No

. coolant pump 2P32D to RCS cold leg (discharge reactor vessel side of RCP 2P32D)

RCS RCS4)01 2CCA-10-30" Cold leg from reactor 15-009 Circumferential weld in No No No No No , No No No coolant pump 2P32D to RCS cold leg next to reactor vessel reactor coolant pump (discharge side of RCP 2P32D)

Desradetson Mechenems T-Hermal Fatigue P - Franary Water Strew Commion Cradung (PWsCC) M-ML-- " ,;- ::j bdluenced Carrossen (MIC) F-Flow AccelerseedCarrossee C-Commion Cracting I-Imersranular strew emos an Cracking 00scr) E - Erosion -Cavitation 0 -Other

'N' N'"'"" " #" A""-oi2. Rn. 00 FMECA - Degradation Mechanisms Page B4 of B36 Weld System ID Segment ,,

Line Number Line Description Number Weld Location T C P I M E F 0 RCS RCS-001 2CCA-10-30" Cold leg from reactor 15-010 Branch connection weld No No No No No No No No coolant pump 2P32D to for Silinein RCS cold leg reactor vessel (discharge side of RCP 2P32D)

RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-001 Weld at RCS cold leg for No No No No No No No No line from RCS cold leg Pressurizer mais spray line 2P32A RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-002 Upstream ofelbow #6 No No No No No No No No line from RCS cold leg 2P32A RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-003 Domistream of elbow #6 No No No No No No No No line from RCS cold leg 2P32A RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-004 Upstream ofcibow #4 No No No No No No No No line from RCS cold leg 2P32A RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-005 Downstream ofelbow #4 No No No No No No No No line from RCS cold leg 2P32A RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-006 Upstream ofelbow #5 No b No No No No No No line from RCS cold leg 2P32A RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-007 Downstream of elbow #5 No No No No No No No No line from RCS coldleg ,

2P32A Deeradstion Mechannms T-umnal Fatigue P - Prunary Water Stress Common Cracbng C WSCC) M - Micretnologicaffy InGuenced Cerrosace (MIC) F- Flow Accelerated Cerroman C- Corremon Cracking I- Intergranular Stress Carroman Crachng (IOs CC) E- Erosion -Cavitauen 0 -Other O O O

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FMECA - Degradation Mechanisms C"'"'"

" d " Y j $ # ,

W eld System ID Segment Line Number Line Description Number Weld IAcation T C P I M E F 0 RCS RCS-001 2CCA-13-3" Pressurizer main spray 27-040 Downstream of piping No No No No No No No No

' line fmm RCS cold leg segment #18 2P32A RCS RCS-001 ?CCA-13-3" Pressurizer main spray 27-041 Upstream ofelbow #17 No No No No No No No No line from RCS cold le-2P32A RCS RCS-001 2CCA-l?-3" Pressurizer main spray 27-042 Downstream ofpiping No No No No No No No No line from RCS cold leg segment #3 (ISO 2CCA-2P32A 13-1) .

RCS RCS-001 2CCA-14-3" Pressurizer main spray 26-001 Weld at RCS coldleg for No No No No No No No No line from RCS coldleg pressurizer main spray line 2P32B RCS RCS-001 2CCA-14-3" Pressurizer main spray 26-001 A Upstream orelbow #5 No No No No No No No No line from RCS cold leg 2P32B RCS RCS-001 2CCA-14-3" Pressurizer main spray 26-003 Downstream of cibow #5 No No No No No No No No line from RCS cold leg 2P32B RCS RCS-001 2CCA-14-3* Pressurizer main spray 26-004 Upstream ofelbow #6 No No No No No No No No line from RCS cold leg 2P328 RCS RCS-001 2CCA-14-3" Pressurizer main spray 26-005 Downstream ofelbow #6 No No No No No No No No line from RCS cold leg 2P32B Desradmuan Mcassusens T-Thmani Fatigue P - Prvnary Water Stress Cerroman Cr4 (P%W M - MicrobiolopesNy inDeenced Cwessan (MIC) F-Flow AcceleratedCorremen C-Carmuen Craciung I-Interpanetar Stress Cerroman Cracbng OGs CC) E - Eremien -Cavnetson 0- Other

f

'*" FMECA - Degradation Mechanisms N""""" #" *""#4 #" #8 ,

Page B6 of B36 W eld System ID Segment Line Nun,ber Line Description Number Weld Location T C P I M E F 0 RCS RCS-001 2CCA-14-3" Pressurizer main spray 26-006 Upstream of cibow #4 No No No No No No No No line from RCS cold leg 2P32B RCS RCS-001 2CCA-14-3" Pressurizer main spray 26-007 Donstream ofelbow #4 No No No No No No No No line from RCS cold leg 2P32B RCS RCS-001 2CCA-14-3" Pressurizer main spray 26-036 Between cIbow #5 and No No No No No No No No ,

line from RCS cold leg sockolet #7 2P32B RCS RCS-001 2CCA-14-3" Dressurizer main spray 26-037 Between weldolet #8 and No No No No No No No No line from RCS cold leg cibow #30 2P32B RCS RCS-001 2CCA-14-3" Pressurizer main sprav 26-038 Upstream of elbon #11 No No No No No No No No line from RCS cold leg 2P32B RCS RCS-001 2CCA-15-3" Pressurizer main spray 26-010 Downstream ofercow #2 No No No No No No No No line RCS RCS-001 2CCA-15-3" Pressurizer main spray 26-011 Upstream ofelbow #4 No No No No No No No No line RCS RCS-001 2CCA-15-3" Pressuric. .nain spray 26-012 Downstream orelbow #4 No No No No No No No No line RCS RCS-001 2CCA-15-3" Pressurizer main spray 26-013 Upstream of elbow #6 No No No No No No No No line RCS RCS-001 2CCA-15-3" Pressurizer main spray 26-014 Downstream ofelbow #6 No No No No No No No No >

line Deermistam Medardsms T-Thermal Fatigue P - Pnmary Water Stress Corrosion Crackmg (PWSCC) M-M~ MA*:y Influenced Cerramen (MIC) F-Flow AccelerasedCerm%n C -Corrosen Cracking I - i.e y. A Stress Corrosion CracLing (IGSCC) E- Eremen -Cavitation 0 - Other O O O

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'" FMECA - Degradation Mec:nnisins N'"" sommai2. Rm 00 Page B7 of B36 wcw System ID Sepnent Line Nuneber IJee Descriptica Number WeM Iscaties T C P I M E F 0 <

RCS RCS-001 2CCA 15-3* Pressurizer main spray 26-015 Upstream orelbow #9 No No No No No No No No I line RCS RCS401 2CCA-15-3* Pressurizer main spray 26-016 Downstream ofcibow #9 No No No No No No No No i line RCS RCS-001 2CCA-15-3* Pressurizer main spray 26-017 Upstream orcibow #Il No No No No No No No No I line !

RCS RCS-001 2CCA-15-3* Fressurizer main spray 26 018 Downstream ofelbow fll No No No No No No No No line ,

i RCS RCS-001 2CCA-15-3* Pressunzer main spray 26-019 Upstream ofelbow #13 No No No No No No No No ;

line i i

RCS RCS-001 2CCA-15-3* Pressunzer main spray 26-020 Downstream ofelbow fl3 No No No No No No No No hne l

RCS RCS40I 2CCA-15-3* Pressunzer main spray 26-020A Between piping e 6--- --- No No No No No No No No i line #26 and #14 i RCS RCS-001 2CCA-15-3* Pressunzer main spray 26-039 Downstream ofelbow No No No No No No No No i line il1(shown on ISO 2CCA-15-2)

RCS RCS-001 2CCA 15-3* Pressurizer main spray 26-040 Upstream ofelbow #2 No No No No No No No No [

line l RCS RCS-001 2CCA-15-3* Lu kcr main spray 26-041 Upstream ofelber,815 No No No No No No No No ;

line RCS RCS-001 2CCA-15-3* Pressurizer main spray 26-042 Downstream ofelbow sI5 No No No No No No No No line i

Desadutaan W T "IherenalFatigue F- Prunsry Water Stress Carremen Croch mg (PWSCC) M - MicrdneingsceNy Indhsenced Correuse (MIC) F-Fleur AccelerusedCervousse {

C-Carecasan Cracdung 1 "a ,, '--Sereus Csereseen Crack % OGSCC) E- Eressen-Consensen O-Other ;

i i

m IN'7 -

Calculancer No. A-PENG-CALC-012. Rev. 00 FMECA - Degradation Mechan sms i p,,, 33 ,f 33c W eld System ID Segment Line Number Line Denription Number Weld Loestion T C P I M E F 0 RCS RCS-001 2CCA-15-3* Pressunzer main spray 26-047 N .smu of 4" x 3* No No No No No No No No line redixer #20 RCS RCS-001 2CCA-15-3* Pressurizer main spray 26448 Upstream of motor- No No No No No No No No line operated waht 2CV-4654 RCS RCS-001 2CCA-15-3* Pressurizer main spray 26-049 N .swo of motor- No No No No No No No No line opersed wthe 2CV-4654 RCS RCS-001 2CCA-15-3* Pressurizer main sprzy 26-050 +)nream ofelbow s24 No No No No No No No No line RCS RCS-001 2CCA-15-3* T.u>umu main spray 26-051 Dominstream orelbow s24 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressuruer main spray 26-052 t*pstream ofelbow s30 No No No No No No No No line (shown on ISO 2C\ A 1)

RCS RCS-001 2CCA-15-3* Pressurizer nnin spray 27-010 Downstram of elbow s2 No No No No No No No No line RCS RCS-001 2CCA-15-3* P.u>u Lu main spray 27-011 Upstream ofcIbow #4 No No No No Na M No No line RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-012 Amswo of elbow #4 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressuruer main spray 27-013 Upstream ofelbow F7 No No No No No ' No No No iine RCS RCS-001 2CCA-15-3" Pressuruer main spray 27-014 Ams-u ofelbow #7 No No No No No No No No line Iksradmuon Mechernsrns T-umnal Fatigue P - Pnmary Wahr Stress Common Cruieg (PRM M - MmeboolaycaHv bdleesiced Carvenen (MK') F-f%w AccelenedCommen c-Common Cruims t - beerranular Stress Commen Craimg (IGSCC) E-Ereman-Caesaham O-Oswr O 9 O

1M O O O

'# FMECA - Degradation Mechanisnes %'e A~o. AMAMWII Rn. 00 Page B9 of B36 Weld Systein ID Segneemt use Neueber Line Description Number Weld IAcation T C P I M E F 0 RCS RCS-001 2CCA-15-3* Pressurizer nuin spray 27-043 Upstream ofelbow #2 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 27-044 6. ohm ofelbow #17 No No No No No W No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 27-047 Dmmream of 4* x 3* No No No No No No No No line redecer #15 RCS RCS-001 2CCA-15-3* Pressunzer main spray 27-048 Upstream ofelbow fl7 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 27-049 Dominstream ofelbow #17 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 27-050 Opstream ofelbow #19 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 27-051 Downstream ofelbow #19 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-052 Upstream orelbow #21 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 27-053 Dominstream ofelbow s21 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pr:ssunzer main spray 27-054 Upstream of motor- No ~ No No No No No No No line operated wahr 2CV-4656 RCS RCS001 2CCA-15-3* Pressurizer main spray "t7-055 Downstream of motor- No No No No No No No No line operated vahc 2CV-4656 W Medunans T-Vennel Fatigue P - Prunary Water Stress Commen Credung (F%W M-ML* * , ,InnmencedCensamm(MK) F-Fleur AszelarenedCorressem c-Cerrames credung I - :.s ., sens Cerroene Creding OGSCC) E- Eressen -Cavemews 0-Oswr L._.----____

i % 97 C"k*iarm No A-PEYG-CALC-C/2, Rev. 00 FMECA - Degradation Mechanisms Page BIO of B36

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W eld System ID Segment Line Number Line Description Nember Weld location T C P I M E F O RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-056 Upstream ofelbow f25 % No No No No No No No line RCS RCS-001 2CCA-15-3" Pressurizer main spray 27 057 Dom 1tstream of elbow f2* No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressurtzer main spray 27-058 Upsm:am ore! bow #27 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-059 Domitstream of elbow s27 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-060 Upstream of main spray No No No No No No No No line valw 2CV-1651 RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-061 Don 1tstream of main spray No No No No No No No No i line valve 2CV-465I RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-062 Upstream ofelbow #41 No Ne No No No No No No line RCS RCS-001 2CCA-15-3* Piumizer main spray 27-063 Dominstream ofcibow #4i No No No No No No No No line RCS RCS-001 2CCA-15-3* Piu.au uxi main spray 27-064 Upstream of motor- No No No No No No No No line operated valve 2CV-4655 RCS RCS-001 2CCA-15-3* Fiu>usisi main spray 27-065 Dv uwmo of motor- No No No No No No No No line operated valm 2CV-4655 RCS RCS-001 2CCA-15-3* Pressurizer main spray 27-066 Upstream of 3* tee #20 No No No No No No No No line (shown on ISO 2CCA-15-1 Dearadamm Mechanners T nermalFatigue P - Pnmary Water Stress Common Cmimg (F%W M- N Inomenced Cerremsam(MIC) F-floor AccelerusedCarrouwe C-Cem=an Cmies 1 - beergranmier sirem Com== Crackmg CGsCC) E- Ennen-Cavemaan 0-06er O O O

O t )

'N" "'"##"'" ^'" *"##""#

FMECA- Degradation Mechanisms Page Bli of B36 .

W eld System ID Segneemt use Number Une Description Neueber Weld tacaties T C P I M E F 0 RCS RCS-001 2CCA-15-3* Pressurizer main spray 28-041 Downstream of 3* tee #20 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 28-042 Downstream ofelbow f21 No No No No No No No No line and upstream of 3* tee #20 RCS RCS-001 2CCA-15-3* Pressuruer main sprav 28-043 LW orcibow f21 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressurizer main spr.y 28-044 Dv ism.i of elbow #23 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressurizer main spray 28-045 Upstream orelbow #23 No No No No No No No No line RCS RCS M 2CCA-15-?" Pressunzer main spray 28-046 Dv xsm ofmotor- No No No No No No No No line operated vahr 2CV-4653 RCS RCS-001 2CCA-15-3* Pressurizer main spray 28-047 Upstream of motor- No No No No No No No No line operated ulve 2CV-4653 RCS RCS-001 2CCA-15-3* Pressunzer main spray 28-048 Dv &mi.ofelbow f27 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressurizer main spray 28-049 Upstream ofcibow #27 No No No No No No No No line RCS RCS-001 2CCA-15-3* Pressunzer main spray 28-050 Dumism orpressuruer No No No No No No No No line main spray valve 2CV-4652 RCS RCS-001 2CCA-15-3* Pressunzer main spray 28-051 Upstream of y m-un No No No No No No No No line main spray vahe 2CV-4652 Desradshan W._a T-T1.mnal Fanigue F- Prunary Weser Stress Corressen Cracbre (r%W M-M_Z *. . ll1homenceaf Carnn.es(h0C) F-Flow Accelensed Cerromen C-Carro Cr. dung -Inney i., seres.Corre Cradtag(1GsCC) E - Enm.an- Canamesas 0 -Oil.er

'" FMECA - Degradation Mechanisms Caladanon Ah AMMW12 Rm M Pay B12 of B36 Weld System ID Segment Line Number Line Description Number Weld incation T C P I M E F 0 RCS RCS-001 2CCA-15-8" Pressurizer main spray 26-043 Dumouwm of 4* x 3* No No No No No No No No 4"line reducer #16 RCS RCS-001 2CCA-15-4* Pressurizer main spray 26-044 Upstream ofelbow fl3 No No No No No No No No 4" line RCS RCS-001 2CCA-15-4* Pressurizer main spray 26445 Donmstream ofelbow #18 No No tw No No No No No 4*Iine RCS RCS-001 2CCA-15-4* Pressurizer main spray 26446 LW of 4* x 3* No N No No No No No No 4"line reducer #20 RCS RCS-001 2CCA-15-4" Pressuriier main spray 27-015 Downstream of 4* x 3" No No No No No No No No 4"line reducer #8 RCS RCS-001 2CCA-15-4* Pressurizer main spray 27-016 Upstream ofelbow fl0 No No No No No No No No 4*line RCS RCS-001 2CCA-15-4" Pressurizer main spray 27-017 Downstream ofelbow slo No No No No No No No No 4*Iine RCS RCS-001 2CCA-15-4" Pressurizer main spray 27-018 Upstream oreIbow #13 No No No No No No No No 4*line RCS RCS-001 2CCA-15-4" Pressurizer main spray 27-019 DumouwTe ofelbow #13 No No No No No No No No 4" line RCS RCS-001 2CCA-15-4* Pressurizer main spray 27-045 Between 4* x3* reducer #8 No No No No No No No No 4"line and elbow #10 RCS RCS-001 2CCA-15-4" Pressurizer main spray 27-046 Upstream of 4" x 3* No No No No No No No No 4"line reducer #15 Dearadatsm Mec!wasms T-Thermal Fangue F - Pnrnary Water stress Cerrosen Crecbng (F%W M - Microbookycally I %enced Ccauseen (MIC) F-Flow AccelerusedCarmeen C-CommenCr cung 1 - beerersmier stress Cerro.ma Cracbng (rGsCC) E- Eroeien-Centsham O-Other O O O

.O O V V f^d

"# FMECA - Degradation Mechanisms N'#"'" * *"##4 R" "'

Page B13 of B36 Weld IJee Nemeber Line Description Nonsber Weld Imtstsee T C P I M E F 0 Systema ID M ;;

2CCA-154* Pressunzer main spray 28-040 lb.s-.. of 4* x 3* No No No No No No No No RCS RCS401 4" line reducer #19 Auxiliary spray line 29-002 Dv-sen ofelbow f33 No No No No No No No No RCS RCS-001 2CCA-16-2*

from 2CV-4824-2 to common spray line Auxiliary spray line 29403 Upstream ofelbow f30 No No No No No No No No RCS RCS-001 2CCA-16-2*

from 2CV-4824-2 to common spray line Auxiliary spray line 29404 Dom 1mstream ofcibow #29 No No No No No No No No RCS RCS-001 2CCA-16-2*

from 2CV4824-2 to common spray line Auxiliary spray line 29-005 Upstream ofc! bow #29 No No No No No No No No RCS RCS-001 2CCA-16-2*

from 2CV-4824-2 to common spray line Auxiliary spray line 29406 Dominstream ofelbow #13 No No No No No No No No RCS RCS-001 2CCA-16-2*

from 2CV-4824-2 to common spray line Auxiliary spray line 29-007 Upstream ofcibow #13 No No No No No No No No RCS RCS-001 2CCA-16-2*

from 2CV-4824-2 to common spray line Auxiliary sp;ay line 29-008 Downstream ofelbow fl4 No No No No No No W W RCS RCS-001 2CCA-16-2*

from 2CV-4824-2 to common spray line Desradmica Medanen.

M-L ' . ,InAmen:edCanonen(nOC)

F-flew Accelermed Carremen T-Thermal Fatigue P - Prunary Waner Serns Commen Crackms (PEW I-irnersran=Isr seras carremen crack =ig(10 SCC) E-Eremen-Cavessmes 0-Odner c-cerramen cracLeis

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'*" FMECA - Degradation Mechanisms C'"" " #" A*"C-## 8" "

Page Bit of B36 Weld System ID Segment Line Number IJae Description Number Weld Imstion T C P I M E F 0 RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-009 Upstream ofelbow #I4 No No No No No No No No from 2CV-4324-2 to common spray line RCS RCS401 2CCA-16-2" Auxiliary spray line 29-010 Donnstream orelbow #15 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-011 Upstream ofelbow fl5 No No No No No No No No from 2CV-4324-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-012 Downstream orelbow fl6 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-013 Upstream of cIbow #16 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-014 Downstream ofelbow #17 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 29-015 Upstream ofelbow #17 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 29416 Dv-um m.. of elbow #18 No No No No No No No No from 2CV-4824-2 to common spray line Desraddson Mecherusms T-nermal Fatigue F - Prvnery Weser stress Cerreeman Crackmg (Pug M - MicrJoslogicmRy infhsenced Carmace (MIC) F-Fleur AccelerusedCamsesme C-Carres an Cracking 1 - beersranular stress Commeon Crackmg (IGSCC) E -Eressen-CavWatsom O-Other O O O

3 t'~'8 (0 pV U

'# # FMECA - Degradation Mechanisms C""'""** *""##1 R" 88 Page Bl3 of B36 W eld System ID Segment Line Number Liec Description Number Weld Location T C P I M E F 0 RCS RCS-001 2CCA-16-2* Auxiliary spray line 29-017 Upstream ofelbow #18 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 29 018 Downstream orcheck No No No No No No No No from 2CV-4824-2 to valve 2CVC-28A common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 29-021 Dominstream orelbow #31 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* Auxiliary spraw line 29-022 Upstream ofelbow #31 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 29-023 Dum m. m of2*tce#24 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* ' Auxiliary spray line 29-024 Upstream of 2* tec #24 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 29-025 Upstream of 2* tec #24 No No No No No No No No from 2CV-4824-2 to common spray line .

RCS RCS-001 2CCA-16-2* Auxiliary spray line 29-042 Upstream ofelbow #1 No No No No No No No No from 2CV-4824-2 to common spray line Desradianos Mh T-Thermal Fatigue P - Pnmary Waeer Stress Comnum Cracbng (P%W M - Micresolaycally Innuenced Carreusen (MIC) F.fleur Accelerused Cerresum C-Cerrossen Cracking I

e ,, Siress Cerroman crecing(IGSCC) E- Eremos -Cavesasag O-Oder l

'# FMECA - Degradation Mechanisms N'*'"*" ^'" **##4 #" "

Page Bl6 of B36 W eld System ID Segment Line Number Line Description Number Weld Location T C P 8 M E F 0 RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-043 Dumohcam of elbow #1 No .4 No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-044 Upstream cfelbow #2 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 294)45 Downstream orcibow f2 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29446 Upwnstream of elbow #3 No No No No No No No No from 2CV-tS24-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-047 Dom wnstream ofelbow No No No No No No No No '

from 2CV-4824-2 to #3 common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-048 Upstream of cibow #4 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29449 Domistream ofcibow #4 No No No No No No No No from 2CV-4824-2 to common spray line RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-050 Upstream of cibow #5 No No No No No No No No from 2CV-4824-2 to common spray line ,

Dearedsma Mechumans T-Dermal Feigue P - Pnmary Water Stres Corremon Oncing (Pug M - MicretnokycmDy Inneenced Commen ().CC) F-The AccelerusedCorremma C-CemmonOncis2 I- Innersrarmier same Cerroman Oncimg OGSCC) E- Eremen-Cavstatsom O -Other O O O

w i pd

' * ' ' G**'"'"" xa FMECA - Degradation Mechanisins Ammni2. Rn. 00 Page BIT of B36 W eld Systems ID Segament Liec Noenber Line Description Nemeber Weld IAesties T C P I M E F 0 RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-051 Dominstream ofelbow #5 No No No No No No No No from 2CV-4824-2 to >

common spray line RCS RCS-001 2CCA-16-2* Auxiliary spray line 29-052 Upstream ofcibow M No No No No No No No No from 2CV4824-2 to common spray line 7

RCS RCS-001 2CCA-16-2" Auxiliary spray line 29-053 Dv..&-- ofelbow #6 No No No No No No No No from 2CV-4824-2 to t common spray line RCS RCS 001 2CCA-2-42* Ilot leg from reactor 07-001 Ciiu... Lmeal weld in No No No No No No No No !

wssel to steam Imp B of RCS hot leg generator 2E248 du.is- of reactor wssel RCS RCS001 2CCA-2-42" Ilot leg from reactor 07-002 Circumferentsal weld in No No No No No No No No ;

wssel to steam Loop B of RCS hot Icg i generator 2E24B downstream ofreactor l ussel !

i RCS RCS-001 2CCA-2-42* Hot leg from reactor 07-005 Cism.Jumeal weld in No No No No No No No No vessel to steam Imp B of the RCS hotleg generator 2E24B doniistream of the SDCS suction line i RCS RCS-001 2CCA-2-42* Hot leg from reactor 07-008 Circumferential weld in No No No No No No No No vessel to steam Loop B of the RCS hot leg

] generator 2E24B dominstream of SDCS l

section line DesadanniMeammum T-Thermal Fatigue P - Frenary Wseer Swens Commene Crachng (P%W M - Micrdeckycelly hdbenced Cerrasse (MIC) F- Flow N Commene C-Commien Cracking I . :.a.

SeemsCommeonOmdums(10 SCC)

E- Eroman -Centunnen 0-Oeur t

to.ser.97 C"Ic"'a'io" No. A-PDU-Gf f.C-Of 2. Rev. 00 FMECA - Degradation Mechanisms Page BIS of B36 Weld System ID Segment Line Number Line Description Number Weld Imention T C F I M E F O RCS RCS-001 2CCJ 42" Hot leg from reactor 07-009 Circumferential weld in No No No No No No No No vessel to steam Imop B of the RCS hot leg generator 2E24B next to steam generator (es-u>Ur sm of SDCS suction line)

RCS RCS-001 2CCA-2-42* Ilot leg from reactor 07-010 Dranch connection weld No No No No No No No No ;

vessel to steam for shutdown cooling i generator 2E24B suction line in Loop B of the RCS hot leg RCS RCS-001 2CCA-29-2* RCS cold leg 2P32B 364)01 Weld at RCS cold leg for No No No No No No No No drain line drain line RCS RCS-001 2CCA-29-2" RCS cold leg 2P32B 36 002 Upstream ofelbow #3 No No No No No No No No drain line RCS PCS-001 2CCA-29-2" RCS coldleg 2P32B 36-003 Dv-u,Le of elbow #3 No No No No No No No No drain line RCS RCS-001 2CCA-29-2" RCS coldleg 2P32B 36 004 Upstream of manual valw No No No No No No No No drain line 2RC-4B RCS RCS-001 2CCA-3-30" Cold leg from steam 10-001 C;rcumferential wrld in No No No No No No No No generator 2E24A to RCS cold leg next to steam reactor coolant pump generator (suction side of 2P32A RCP 2P32A)

RCS RCS-001 2CCA-3-30* Cold leg from steam 10-002 Circumferential utid in No No No No No No No No generator 2E24A to RCS cc,Id jeg (suction side l

. reactor coolant pump of RCP 2P32A) 2P32A Deu-adsrum Mahneers T-11mrmal Fatigue P- Pnenary Waner Stress Correenn Crackmg (PWSCC) M - ML*le"y Indhsenced Carremenn (M:C) F-flew Accelersmed Carremum C-Cerrosen Crackmg I-Iraergranstar Stress Commen CracLeg(IGSCC) E - Ersion -Cavgassen 0 -Oqher O O O

m s n\ (b \

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'# FMECA - Degradation Mechanisms N"" " h NMWil Rm 00 Page B19 of B36 W eld Systems ID 4: Line Number Line Descripties Noenber Weld I.mestice T C F I M E F 0 RCS RCS-001 ~ A-3 Cold leg from steam 10405 Circumferential wtld iv No No No No No No No No generator 2E24A to RCS cold leg (sucuan side reactor coolant pump of RCP 2P32A) 2P32A RCS RCS-001 2CCA-3-30* Cold leg from steam 10 008 Ciiu Jume.1weldin No No No No No No No No generator 2E24A to RCS cold leg (secuan side reactor coolant pump ofRCP2P32A) 2P32A RCS RCS401 2CCA-3-30* Cold leg from sacam 10411 Circumferentsal utid in No No No No No No No No generator 2E24A to RCS cold leg downstream reactor coolant pump ofIctdown line (suctiru 2P32A side ofRCP 2P32A)

RCS RCS401 2CCA-3-30* Cold leg from steam 10-014 CiiumJuma-J wrld in No No No No No No No No generator 2E24A to RCS cold leg (sucuon side reactor coolant pump of RCP 2P32A) 2P32A RCS RCS401 2CCA-3-30" Cold leg from steam 10-015 Cin Jumest weld in No No No No No No Nd No generator 2E24A to RCS cold leg next to reactor coolant pump reactor coolant pump 2P32A '

(suchon side ofRCP 2P32A)

RCS RCS-001 2CCA-3-30" Cold leg from steam 10-016 Branch connecuan weld No No No No No No No No generator 2E24A to forletdownlinein RCS reactor coolant pump cold leg (suchon side of 2P32A RCP 2P32A)

RCS RCS-001 2CCA-30-2* RCS coldleg 2P32C 38-001 Weld at RCS cold leg for No No No No No No No No drain line drain line Desradeian W T-Themel Fatigue F- Pnenary Water Steens Carremen Cracksig (r am M - Micreb.olopenity Foluenced Carreuses (MIC) F-flour Acanterused Commsen C-Cerrassan Onding I- beergraader Stres Cerremon Cradmig (1GSCO E- Esesean-C maassa 0-Osbar

iosepm FMECA - Degradation Mechanisms Calc ='ar""r No. A-PDGCALC-012 Rev. 00 Page B:0 of B36 W eld System ID Segment Line Number Line Dewription Number Weld lecation T C P I M E F 0 RCS RCS@l 2CCA-30-2* RCS cold leg 2P32C 38-002 Upstream ofelbow #3 No No No No No No No No drain iiae RCS RCS-001 2CCA-30-2* RCS cold leg 2P32C 38-003 Downstream ofelbow #3 No No No ho No No No No dHn line RCS RCS-001 2CCA-30-2* RCS cold leg 2P32C 38-004 Upstream ormanual vahr No No No No No No No No drain line 2RC-4C RCS RCS@I eCCA-31-2* RCS cold leg 2P32D 39-001 Weld at RCS cold leg No No No No No No No No drain line 2P32D RCS RCS@l 2CCA-31-2* RCS cold leg 2P32D 39-002 Upstream orelbow #3 No No No No No No No No drain line RCS RCS-001 2CCA-31-2* RCS coldleg 2P32D 39-003 Downstream ofelbow #3 No No No No No No No No drain line RCS RCS-001 2CCA-31-2* RCS cold leg 2P32D 39-004 Upgream of manual vaht No No No No No No No No drain line 2RC-4D RCS RTS-001 2CCA-32-2* RCS hot leg to steam 35-002 Upstream ofelbow #7 No No No No No No No No generator 2E24A drain line RCS RCS-001 2CCA-32-2* RCS hot leg to steam 35-003 Du..oumii of elbow s7 No No No No No No No No generator 2E24A drain line RCS RCS4)01 2CCA-32-2* RCS hot leg to steam 3 % 04 Upstream ofcibow #8 No No No No No No No No generator 2E24A drain line Dearadaten Mecherusers T-Therrnal Fatigee P - Prunary Water Stress Cerrasson Cracking (FWSCC)

C-Carroman Cracking M-E l LA IndeencedCarremen(MIC) F-flour AccelerenedCerronen I-Intergrannter Stress Cerronen Cracbng OGSCC) E- Erweren -Caventice 0- Other O O O

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'" FMECA - Degradation Mechanisms N#"#"*'" A~a AMMC-012. Rm W l' age B21 of B36 W eld Systen ID Segnse:it Line Number Line Descripties Noneher Weld Location T C P I M E F O RCS RCS-001 2CCA-32-2* RCS hot legio steam 35-005 Du &-iofelbow #8 No No No No No No No No generator 2E24A drain line RCS RCS 001 2CCA-3r-2* RCS hot leg to steam 35-006 Upstream ofcibow #9 No No No No No No No No generator 2E24A drain i line RCS RCS M 2CCA-32-2* RCS hot leg to steam 35-007 Dv-n-i of chow #9 No No No No No No No No generator 2E24A drain line RCS RCS-001 2CCA-32-2* RCS hot leg to steam 35-008 Upstream ofcibow #10 No No No No No No No No generator 2E24A drain line RCS RCS-001 2CCA-32-2* RCS hot leg to steam 35-009 Downstream ofcibow #10 No No No No No No No No generator 2E24A drain j line ,

RCS RCS-001 2CCA-32-2* RCS hot leg to steam 35-010 Upstream 6 -.' a #11 No No No No No No No No generator 2E24A drain line RCS RCS-001 2CCA-32-2* RCS hot leg to steam 35-011 Dv-a- of 2* tec #11 No No No No No No No No generator 2E24A drain line RCS RCS-001 2CCA-32-2* RCS hot Icg to steam 35412 Upstream of manual valve No No No No No No No No I generator 2E24A drain 2RC-4E line t

Desradmanen Meden===

T-nermal Fatigue P - Primary Waser Stress Cerveneen Crackms (F%W hi- E J _ _ Z, tedimenaced Camusam (MIC) F-Flow Accelerened Careemen C-Carrossen CancMng I - Ireeryweler serens Cerresem Crackmg (1GsCC) E- Eresson-Canesesse 0-Oeier 6

i

FMECA - Degradation Mechanisms C"'"'"' #" NN## " " "

Page B22 of B36 Weld System ID Segment Line Nember Line Description Number Weld Location T C P I M E F O RCS RCS-001 2CCA-32-2" RCS hot leg to steam 35-015 Dmmstream of 2" tee No No No No No No No No generator 2E24A drain #11(refueling lesel side) line RCS RCS-001 2CCA4-30" Cold leg from reactor II-001 Circumferential wrld in No No No No No No No No coolant pump 2P32A to RCS cold leg next to reactor vessel reactor vessel (discharge <

side of RCP 2P32A)

RCS RCS-001 2CCA4-30" Cold leg from reactor II-002 Circumferential wrld in No No No No No No No No coolant pump 2P32A to RCS cold leg duouhn reactor vessel of SIline (discharge side of RCP 2P32A)

RCS RCS-001 2CCA4-30' Cold leg from reactor II-005 Circumferentzal wrid in No No No No No No No No coolant pump 2P32A to RCS cold leg dowitstream reactor srssel of SIline(discharge side of RCP 2P32A)

RCS RCS-001 2CCA4-30" Cold leg froca rcactor 11-008 Circumferential weld in No No No No No No No No coolant pump 2P32A to RCS cold leg (discharge reactor sessel side of RCP 2P32A)

RCS RCS-001 2CCA4-30" Cold leg from reactor 11-009 Circumferential weld in No No No No No No No No ,

coolant pump 2P32A to RCS cold leg next to reactor vessel reactor coolant pump (discharge side of RCP 2P32A)

RCS RCS-001 2CCA4-30" Cold leg from reactor II-010 Branch connection weld No No No No No No No No coolant pump 2P32A to for SIline in RCS cold leg reactor vessel (discharge side of RCP 2P32A)

Dearadation Mechannms T-Thermal Fangue P - Pnmary Waser Stresa Carrossen Cracksng (PWSCC) M - MR M'y Inf5aenced Correman(MIC) F-Fle= Accelerated Comnica C-Carromen Crackmg 1 - Intsgranular Stresa Cerrosion Cracing (IGSCC) E- Ereason- Cavention 0-Other O O 9

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'N" Calculation No. A-PENG-C4LC-012. Rev. 00 FMECA - Degradation Mechanisms roge s23 of s36 WeW Systene ID Segment line Number Line Descripties Number Weld Emestion T C F I M E F 0 RCS RCS-001 2CCA-4-30" Cold leg from reactor 11-011 Branch connecnon weld No No No No No No No No coolant pump 2P32A ta for pressmuer :: pray line reactor vessel (dtscharge side of RCP 2P32A)

RCS RCS-001 2CCA-5-30* Cold leg from steam 08-001 Cimu Juusalweldin No No No No No No No No generator 2E24A to RCS cold leg next to steam reactor coolant pump generator (sucuon of RCP 2P32B 2P32B)

RCS RCS-001 2CCA-5-30* Cold leg from steam 08-002 Cinu Juusal weld in No No No No No No No No generator 2E24A to RCS cold leg (suchon side reactor coolant pump of RCP 2P32B) .

2P32B RCS RCS-001 2CCA-5-30" Cold leg from steam 08-005 Cisu .Juuaialweldin No No No No No No No No generator 2E24A to RCS coldleg(secuen side reactor coolant pump of RCP 2P32B) 2P32B RCS RCS-001 2CCA-5-30" Cold leg from steam 08-008 CL-.Juuaial weld in No No No No No No No No generator 2E24A to RCS cold leg (sucuan side reactor coolant pump of RCP 2P328) 2P32B RCS RCS-001 2CCA-5-30" Colf leg from steam 08-011 Cinu Juus.Iweldin No No No No No No No No genstor 2E24A to RCS cold leg (secuan side rertor coolant pump of RCP 2P32B) 2P32B l Desadawn Mecb====v

! T-Therumi Fatigue F - Prenary waaer Seems Carremen Craciung (PWsCC) M - Microbielepcmily Innuenced Carremen (MIC) T-flow AcceleraardCorremen c-Corremen Crecung I-IrmerymederseemcerromenCn chngOGSCC) E-Demon-Covenann 0-Oemr

inepm FMECA - Degradation Mechanisms C"Ic' dado"No. A-PEM-CALC-012 Rev. 00 Faxe B24 of B36 Weld System ID Segment Line Number Line Description Number Weld imention T C F I M E F O RCS RCS-001 2CCA-5-30* Cold leg fro.n sicam 08-014 Circumferential utid in No No No No No No No No generator 2E24A to RCS cold leg (sucten side reactor coolant pump of RCP 2P23B) 2P32B RCS RCS-001 2CCA-5-30" Coldleg from steam 08-015 Circumferentialutidin No No No No No No No No generator 2E24A to RCS cold leg next to reactor coolant pump reactorcoolant pump 2P32B (suction side of RCP 2P32B)

RCS RCS-001 2CCA-5-30" Cold leg from steam 08-016 Branch connection wrid No No No No No No No No generator 2E24A to for RCS cold led drain line reactor coolant pump (suction side of RCP 2P32B 2P32B)

RCS RCS-001 2CCA-6-30* Cold leg from reactor 09-001 Circumferentral wrid in No No No No No No No No coolant pump 2P32B to RCS cold leg next to reactor srssel reactor vess-J (M.,p side of RCP 2P328)

RCS RCS401 2CCA-6-30* Cold leg from reactor 09-002 Circumferential utId in No No No No No No No No coolant pump 2P32B to RCS cold leg (discharge reactor vessel side of RCP 2P32B)

RCS RCS-00I 2CCA-6-30* Cold leg from reactor 09405 Circumferential weld in No No No No No No No No coolant pump 2P32B to RCS cold leg i, .h..

reactor vessel ofcharging nozzel (discharge side of RCP 2P32B)

IL-f tion MA -.

T-Thermat Fatigue P - Pnmary Water Stress Cerressen Craclung (PWSCC) y M-Micrdnedaycap bilmenced CorreesenOL81C) F-rieur AcurlerusedCarreeuwe C Carreusen CracEng I-:.a.y- : Stress Cceressen Crackang(IGSCC) E - Ereman -Caviestaan 0 -Other e 9 _- _

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FMECA - Degradation Mechanisnes fj#(( ,

W eld Systems ID Segneemt Line Noseber Line Descripties Number Weld Leesties T C P I M E F 0 RCS RCS401 2CCA430* Cold leg from reactor 09-008 Cirannferenhal weld in No No No Na No No Ne No coolant pump 2P32B e RCS cold leg (dsd, '

reactor vessel side of RCP 2P32B)

RCS RCS-001 2CCA +30" Cold leg from reactor 09-009 Ci w..Juu.ii.I weld in No No No No No No No No coolant pump 2P328 to RCS cold leg next to reactorsrssel reactor coolant punp

, (discharge side of RCP 2P32B)

RCS RCS-001 2CCA4-30" Cold leg from reacter 09410 Branch conaccuan weld in No No No No No No No No ,

coolant pump 2P32B to RCS coldleg for charging '

reactor sessel nozzle (discharge side of RCP 2P32B)

RCS RCS-001 2CCA4-30* Cold leg from reactor 0941I Branch connection weld in No No No No No No No No ;

coolant pump 2P32B to RCS cold leg ror Si nozzle reactor srssel (discharge side of RCP 2P32B)

RCS RCS-001 2CCA4-30" Cold leg from reactor 09-012 Branch connection weld in No No No No No No No No coolant pump 2P32B to RCS coldleg for reactor srssel v.-.La sprayline (discharge side of RCP 2P32B)

RCS RCS-001 2CCA-7-30* Cold leg from steam 12-001 Cimm J u.iial weld in No No No No No No No No generator 2E24B to RCS cold leg next to steam reactor coolant pump generator (sucuan side of

, 2P32C RCP 2P32C) asst =dem M=f=ai==

T-Tiermal Fatigue F - Pnmary Waner Seen Cerrimen Creciung (PRW M - Macrednalossemity inomenced Cerressan (MIC) F-Flein AccelermaedCarremen t c-Cone-onCr chng 1 - :.s ., s=== C. r, on Cr chng(1GSCC) E- Eressee -Cavemasem O -Oder

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A * " # '; # " 88 FMECA - Degradation Mechanisms Page B:6 of B36 W eld System ID Segment IJee Number Line Description Number Weld Imcation T C P I 31 E F 0 RCS RCS40I 2CCA-7-30* Celd leg from steam 12-002 Circumferential weld in No No No No No No No No generator 2E24B to RCS cold leg (suction side reactor coolant pump of RCP 2P32C) 2P32C RCS RCS401 2CCA-7-30" Cold leg from steam 12-005 Circumfertritsal weld in No No No No No No No No g-nerator 2E24B to RCS cold leg (sucten side reactor coolant pump of RCP 2P32Q 2P32C RCS RCS-001 2CCA-7-30* t'old leg from steam 12 008 Chuunfuudal weld in No No No No No No No No generator 2E24B to RCS cold leg upstream o' reactor coolant pump drain line (suction side-2i'32C RCP 2P32Q RCS RCS-001 2CCA-7-30* Cold leg from steam 12-0II Circumferentnl weld in No No No No No No No No generator 2E24B to RCS cold Icg dumsr uu reactor coolant pump ofdrain line (sucten side 2P32C of RCP 2P32C)

RCS RCS-001 2CCA-7-30" Cold leg from steam 12 014 Circumferential wrid in No No No No No No No No generator 2E24B to RCS cold leg (sucten side reactor coolant pump of RCP 2P32C) 2P32C RCS RCS-001 2CCA-7-30" Cold leg from steara 12-015 Circumferential neld in No No No No No No No No generator 2E24B m RCS cold leg r m to reactor coolant pcmp reactorcoolant pump 2P32C (sucten side of RCP 2P32C) pe.raamioo w w ,.n.

T-Thermet Fatigue P - Prenery Weser Stres Common Crackmq (P%W M -M_22,, :y kdleenced Commaan(1 LOC) F-Flow AccelerseedCerrousan

c. corr mencreas, t-% -sien.com oncr=Angposec; E- Erensin -Cavemiiise 0- N O _

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'" FMECA- Degradation Mechanisms C"k"#" No a-PEm-c4tc-oi2. Rcr. 00 Page B27 of B36 WeW System ID Segneent Line Noseber IJee Description Nuneber Weld Imaties T C P I M E F 0 RCS R.CS-001 2CCA-7-30* Cold leg from steam 12-016 Branch connection weld No No No No No No No No generator 2E24B to for RCS cold leg drain line reactor coolant pump (suction side of RCP 2P32C 2P32Q RCS RCS-001 2CCA-8-30* Cold leg from reactor 13-001 Cin 'uudial weldin No No No No No No No No coolant pump 2P32C to RCS cold leg next to reactor vessel reactorwssel(drscharge !

, sideofRCP2P32Q RCS RCS-001 2CCA-8-30* Cold leg from reactor 134102 Cimanferential weld in No No No No No No No W coolant pump 2P32C to RCS cnid leg

mm ,

reactor vessel ofcharging line (discharge i side of RCP2P32Q RCS RCS-001 2CCA-8-30" Cold leg from reactor 13-005 Circumferentsal weld in No No No No No No No No i coolant pump 2P32C to RCS cold leg downstream rescest wssel ofcharging line (discharge !

side of RCP 2P32Q ,

RCS RCS-001 2CCA-8-30" Cold leg from reactor 134X)8 Circumferentsal weld in No No No No No No No No l coolant pump 2P32C to RCS cold leg (discharge l reactor vessel side of RCP 2P32O RCS RCS-001 2CCA-8-30* Cold leg from reactor 13-009 Circumferentsal weld in No No No No No No No No '

coolant pump 2P32C to RCF cold leg next to reactor wssel reactor coolant pump .

(discharge side of RCP 2P32Q

. I Desradmanni W T-Thennet retene r-Frunarywesersim.c c,=ams(ruxr) u - w u , ,.gi ac (wic) 7. rio. 4,,,i,,,,,4 c, ,,,

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'" C"midmn A'a AMRAWI R m 00 FMECA - Degradation Mechanisms Page B:8 of B36 W eld System ID Segment Une Number Line Description Number Weld Location T C I P M E F O RCS RCS401 2LCA-8-30* Cold leg from reactor 13-010 Branch connection meld No No No No No No No No coolant pump 2P32C to for charging line in RCS reactor sessel cold leg (drscharge side of RCP 2P32C)

RCS RCS-001 2CCA-8-30* Cold leg from reactor 13-011 Branch weld for SIline in No No No No No No No No coolant pump 2P32C to RCS cold leg (Judege reactor vessel side of RCP 2P32C)

RCS RCS001 2CCA-9-30' Cold leg from steam 14-001 Circumferential weld in No No No No No No No No generator 2E24B to RCS cold leg next to steam reactor coolant pump generator (sucten side of 2P32D RCP 2P32D)

RCS RCS-001 2CCA-9-30' Cold leg from steam 14-002 Circumferential weld in No No No No No No No No generator 2E28** to RCS cold leg (sucten side reactor coolant ptmp of RCP 2P32D) 2P32D RCS RCS-001 2CCA-9-30* Cold leg from steam 14-005 Ciiuuufuudial weld in No No No No No No No No generator 2E24B to RCS cold leg (suction side reactor coolant pump of RCP 2P32D) 2P32D RCS RCS-001 2CCA-9-30* Co'd leg from steam 14-008 Circumferen:2al weld in No No No No No No No No gercrator 2E24B to RCS cold leg opumii of scactor coolant pump drain line (sucten side of 2P32D RCP 2P32D)

RCS RCS-001 2CCA-9-30' Told leg from steam 14-011 Circumferentral wrld in No No No No No No No No g:nerator 2E24B to RCS cold leg dui uwm.; '

mtor coolant pump ofdrain line (sucten side 2F32D of RCP 2P32D)

Derradstreo Mechancru T IhermalFatigue P - Prenery Water Stres Correnen Cruclung (P%M M - Maebeoicpcally Iremenced Cerreman (MIC) F-Flow AccelermaedCerremen C-Corressen Oschlag I - :.a. , Sims Cerremon Oncksng(1GSCC) E - Erasecus - Cavenemu 0 -Oder O O O

n n

'" FMECA - Degradation Mechanisms N '"'""*" w m m m i2.R w. m Page B29 of B36 W eld System ID Segment IJee Number time Description Number Weld I.mcaties T C P I M E F O RCS RCS-001 2CCA-9-30" Cold kg from steam 14414 Circumferential weld in No No No No No No No No generator 2E248 to RCS cold leg (suction side reactor coolant pump of RCP 2P32D) 2P32D RCS RCS-001 2CCA-9-30" Cold leg from steam 14-015 Circumferential weld in No No No No No No No No generator 2E24B to RCS cold Icg next to reactor coolant pump reactor coolant pump 2P32D (suction side of RCP 2P32D)

RCS RCS-001 2CCA-9-30* Cold leg from steam 14-016 Branch connection weld Nc No No No No No No No generator 2E24B to for drain line in RCS cold reactor coolant pump leg (suction side of RCP 2P32D 2P32D)

RCS RCS-002 2BCA-1-12" Pressurizer surge line 16-00I Pressunzer surge line Ycs No No No No No !<o No Mng weld to RCS het leg RCS RCS-002 2BCA-t-12" Pressudzer surge line 16-002 Upstream of piping Yes No No No No No No No segment #1 RCS RCS-002 2BCA-1-12* Pressurizer surge line 16-003 Upstream ofelbow #1 Yes No No No No No No No l

RCS RCS402 2BCA-1-12" Pressunzer surge line 16-004 Downstream ofcibow #I Yes No No No No No No No RCS RCS-002 2BCA-I-12* Pressunzer surge line 16-005 Upstream ofcibow #2 Yes No No No No No No No RCS RCS-002 2BCA-1-12" Pressunzer sarge line 16 =106 Dv-i=.m ofc!!xm #2 Yes No No No No No Ne No RCS RCS-002 2BCA-I-12" Pressuriier surge line 16-007 Upstream ofcibow #3 Yes No No No No No No No Dearadmason h%

T-Thenal rsaigne P - Prenary "Nuer Stress Carressen Chdi.g (P%1Kr) M-M-. A* . dtyIn8mencedCommeon(MIC) F-flow AccelerenedCerrousen c-carroman crucians I-insersr=== tar Stress Cmessen Crading OGSCQ E-Erossen Cavitataam 0-Oswr

i4-sy m Cala'dma h A-MGCALC-012. Rn. 00 FMECA - Degradation Mechanisms Page B30 of B36 Weld System ID Segment Line Number Line Description Number Weld leestion T C P I M E F 0

.RCS RCS-002 2BCA-I-12" Pressurizer surge line 16408 Downstream ofcibov 33 Yes No No No No No No Na RCS RCS-002 2BCA-1-12" Pressurizer surge line 16 009 Upstream cfelbow #4 Yes No No No No No No No RCS RCS402 2BCA-t-12" Pressurir r surge line 16-010 Dvomi.wm of elbow f4 Yes No No No No No No No RCS PCS402 2BCA-t-12* Pressurizer smge line 16-011 Upstream ofcibow #5 Yes No No No No No No No RCS RCS402 2BCA-t-12' Pressurtzer surge line 16-012 Duw misr.am of elbow s5 Yes No No No No No No No RCS RCS-002 2BCA-I-12* Pressurizer surge line 16-013 Douwstream orpiping Yes No No No No No No No segment #6 RCS RCS-002 2BCA-t-12" Pressurizer surge line 16-014 F.u-uu surge line Yes No No No No No No No connecting wrld to pressurizer RCS RCS-002 2BCA-14-4" Pressurms vent line - 43-025 Upstream of motor- Yes No No No No No No No 4" sectice operated vahr 2CV-4730-1 RCS RCS-002 2BCA-I4-4* Pressurizer wnt line - 43-029 Donastream of 6* x 4* Yes No No No No No No No 4" section reducer #30 RCS RCS-002 2BCA-144* T.u aci vent line 42-001 Weld at nozzle for Yes No No No No No ~4 No '

pressurizerunt/LTOP line RCS RCS-002 2BCA-144* F u ku ventline 43-019 Weld at y.u-uu flange Yes No No No No No No No

2 RCS RCS-002 2BCA-144* F,u-uct unt line 43-020 Upstream of 6* tee #5 Ys No No No No No No No RCS - RCS-002 2BCA-144* Fiu ku wnt line 43-022 Downstream of 6* tee #5 Yes No No No No No No No RCS RCS-002 2BCA-144" 7.u kci vent line 43-023 Upstre m of 6* tee #34 Yes No No Ne No No Na No Desreameien Mechare-ns T-Therwat ras. gee r- Pnmary waier sire a Common Cruisis (rwscc) u - us:rab.asos,enny bdimenced Corre-en (ux-) F-rio. Acceler sed Car,wo.

C-Ccercesan Cradmg I- beergrarmlar Stress Commen Crackmg OGSCC) E -Essesa. -Cantmanen 0 -Oiher O O O

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' N'7 FMECA - Degradation Mechanisnes O'"""**

j#[g"[ !

W eld Systene ID Segneemt Line No;mber Line Descriptier Namober Weld IAcation C T P I M E F 0 1

RCS RCS-002 2BCA-14-6* Pressurizer vent line 43-024 Upstream of 6* x 4* Yes No No No No No No No redu:crir8 RCS RCS-002 2BCA-I4-6* Pressunzer vent line 43428 Upstream of 6* x 4* Yes No No No No No No No reducer #3G RCS RCS-002 2CCA-154* Pressurizer main spiay 28408 Du-.s- cf eWw #16 Yes No No No No No No No 4*line RCS RCS-002 2CCA-15-4' Pressurizer main spray 28-009 Upstream of cWw #14 No No Yes No No No No No 4"line RCS RCS-002 . 2CCA-154" F.u>maa main spray 28-010 Dowtistream ofelbow #14 Yes No No No No No No No 4*line RCS RCS402 2CCA-154" Pressurizer main spray 28-011 Bernten cibow #14 and Yes No No No No No No No 4"line cibow #1I RCS RCS402 2CCA-154* Pressurizer main spray 28-012 Upstream ofelbow #11 Yes No No No No No No No 4*line RCS RCS-002 2CCA-154" Pressurizer main spray 28413 Don 1:sucam ofcibow #1I Yes No No No No No No No 4"line RCS RCS-002. 2CCA-154" Pressurizer main spray 28-014 Upstream ofelbow #9 Yes No No No No No No No 4*line RCS RCS-002 2CCA-154" Pressurizer main spray 28-015 Dominstream ofelbow #9 Yes No No No No No No No 4*line RCS RCS-002 2CCA-154" Presurizer main sprn 28-016 Upstream ofelbow #7 Yes No No No No No No No 4*line Dearadetson Mediansess T-Dmnal Fatigue P-PrvneryWaterStr ascerrassenCrackmg(r%W M - Mscro sweseqpenny Inamenced Corremen (nGC) F-Flow AccelerasedCarmosen c-Cerroman creaang . I-Innerar===1=r sir === corre=== creduas (iosCC) E- Eremen-Caveseren 0-Odur

14-sep.97 FMECA - Degradation Mechanisms C"Ic"I"'""r k. A-PDGCALC-012 Rcr. 00 Page B32 of B36 Weld System ID S:gment Une Number Line Description Nember Weld Location T C P I M E F O RCS RCS-002 2CCA-15-4" Pressurizer main spray 28-017 Downstream ofelbow #7 Yes No No No No No No No 4"line RCS RCS-002 2CCA-15-4" Pressurizer main spray 28-018 Updefbow #5 Yes No No No No No No No 4'line

[2CS RCS-002 2CCA-154" Pressurizer main spray 28-019 Dowt. stream of cIbow #5 Yes No No No No No No No 4"line RCS RCS-002 2CCA-l!-4" Pressurizer main spray 28-020 Between piping xy.m.h Yes No No No No No No No 4"line #3 and #46 RCS RCS-002 2CCA-154" Pressurizer main spray 28-021 Upstream ofelbow f2 Yes No No No No No No No 4*line RCS RCS-002 2CCA-15-4" Pressunzer main spsy 28-022 Donnstream ofelbow #2 No Yes No No No No No No 4"line RCS RCS-002 2CCA-15-4* Pressurizer main spray 28-023 Downstream of piping Yes No No No No No No No 4"line segment #1 RCS RCS-002 2CCA-154" Pimmha main spray 28-024 Weld at pressunzer spray Yes No No No No No No No 4"line nozzle RCS RCS-002 2CCA-154" Pressunzer main spray 28-038 Upstream ofchlow #16 Yes No No No No No No No 4*line RCS RCS-002 2CCA-15-4" Pressurizer main spray 28-039 Dumh.. of 4* x 4' x Yes No No No No .No No No 4"line 2" tee #I8 (4* side)

RCS RCS-002 2CCA-16-2* Auxiliary spray line 29-054 Upstream ofcibow #7 Yes No No No No No No No from 2CV-4824-2 to common spray line Ikaradsucn Medsunisms T-Thermal Fatigue P- Primary Water Swess Cerrasson Cracing(P%?T) M - Micresolegicany InSmenced Cerresum NIC) F-Th= AccelerseedCorresum C-Corresean Cracbng 1 -linerperusler stress Cerroserm Ctackmg OGs CC) E - Eronen -Caviasuon 0-Oder O O O

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'*" FMECA - Degradation Mechanisms C""'"**" Ne AMN4/2. Rm 00 l' age B33 of B36 wew System ID Segment Line Number Line Descripties Number Weld teesties T C P I M E F 0 RCS RCS-002 2CCA-16-2* Auxiliary spray line 29455 Downstream ofelbow #7 Yes No No No No No No No from 2CV-4824-2 to common spray hne RCS RCS-002 2CCA-16-2* Auxiliary spray line 29-056 Upstream of 4* x 2* tec Yes No No No No No No No from 2CV-4824-2 to #18 (as shown on ISO common spray line 2CCA-15-1)

RCS RCS-003 2CCA-16-2" Auxiliary spray line 29-019 Upstream orcheck vahr No No No No No No No No from 2CV-4824-2 to 2CVC-28A common spray line RCS RCS-003 2CCA-16-2* Auxiliary spray line 29-020 Dums-s of motor- No No No No No No No No from 2CV-4324-2 to operated wahr 2CV-4824-2 common spray line RCS RCS-003 2CCA-29-2* RCJ cold leg 2P32B 36-005 Downstream of manual No No No No No No No No drain line vahr 2RC-4B RCS RCS-003 2CCA-29-2* RCS coldleg 2P32R 36-006 Upstream of cibow #6 No No No No No No No No drain line RCS RCS-003 2CCA-29-2* RCS coldleg 2P32B 36-007 Doms-n ofcibow #6 No No No No No No No No drain line

~RCS RCS-003 2CCA-29-2* RCS cold leg 2P32B 36-008 Upstream ofelbow #7 No No No No No No No No drain line RCS RCS-003 2CCA-29-2* RCS coldleg 2P32B 36-009 Downstream ofelbow 87 No No No No No No No No drain line RCS RCS-003 2CCA-29-2* RCS coldIcg 2P32B 36 010 Upstream of manual vaht No No No No No No No No drain line 2RC-5B nestadecen F T "DiennelFuzigue P Pnnenry Water Swess Cerrow= Cruik (F%M M-MR . .InnsencedCarremen(MK) F-Flow Acce4erusedCerrones c-Carresum Crack-g - :.e ., swe= Cerroman Crackang(IGsCC) E- Erenam -Cavammen 0-oder t_

10-Sep-97 -

FMECA - Degradation Mechannsnts Cakulatun M. A-PENGCtLC-012. Rev. 00 7,y, 3,, ,f gy,

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Weld System ID Segment Line Number Line Description Number Weld Imcation T C P I M E F 0 RCS RCS403 2CCA-30-2* RCS cold leg 2P32C 38405 Downstream of manual No No No No No No No No drain line valve 2RC-*C RCS RCS-003 2CCA-30-2* RCS cold leg 2P32C 38-006 Upstream ofcIbow #6 No No No No No No No No drain line RCe RCS-003 2CCA-30-2" RCS cold leg 2P32C 38407 Downstream ofcibow #6 No No No No No No No No drain line RCS RCS-003 2CCA-30-2* RCS cold leg 2P32C 38-003 Up:tream ofelbow #7 No No No No No No No No drain line RCS RCS-003 2CCA-30-2* RCS cold leg 2P32C 38409 Downstream ofcIbow #7 No No No No No No No No drain lix RCS RCS-003 2CCA-30-2* RCS coldleg 2P32C 38-010 Upstream of marraal raht No No No No No No No No drain line 2RC-SC RCS RCS-003 2CCA-31-2* RCS cold leg 2P32D 39-003 Downstream of manual No No No No No No No No drain line vahr 2RC-4D RCS RCS-003 2CCA-31-2* RCS cold leg 2P32D 39-006 Upstream cfelboz #6 No No No No No No No No drain line RCS RCS-003 2CCA-31-2* RCS cold leg 2P32D 39-007 Downstream ofcibow #6 No No No No No No No No drain line RCS RCS403 2CCA-31-2* RCS coldleg 2P32D 39-008 Upstream ofcibow #7 No No No No No No No No drain line RCS RCS-003 2CCA-31-2* RCS cold leg 2P32D 39-009 DumhTi of elbow f7 No No No No No No No No drain line Dearudarm Mederusrns T-Theral Feigue P- Pnrnary Weer Stress Carressen Chiing(PEW M - Microbie4ogica!!y Infbenced Caminen (MIC) F-fle= AccrierusedCommere c-Carrossee Crwinne 1 -Ireergranatar sirem Cerresien cradung (rGSCC) E- Eressee-Cmtsesee 0 -Other O O O

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FMECA - Degradation Mech nisms " "#" *^'* A d ^ " " # #'" 88 Page B35 of B36 W eld System ID Segment Line Number Line Description Number Weld Leestion T C P I M E F 0 RCS RCS-003 2CCA-31-2" RCS cold leg 2P32D 39-010 Upstream ormanual vaht No No No No No No No No drain line 2RC-5D RCS RCS-003 2CCA-32-2" RCS hot leg io steam 35-013 Downstream ormanual No No No No No No No No ;

generator 2E24A drain vahr 2RC-4E line RCS RCS-003 2CCA-32-2" RCS hot leg to steam 35-014 Upstream of manual vaht No No No No No No Na No generator 2E24A drain 2RC-5E line RCS RCS-004 2BCA-14-3" Pressurizer vent bypass43-033 Downstream of 4" x 4" x Yes No No No No No No No line 3" tee #6 (3* side)

RCS RCS-004 2BCA-14-3" Pressurizer wnt bypass43-034 Upstream ofelbow #I4 Yes No No No No No No No ,

line RCS RCS-004 2BCA-14-3" Pressurizer vent Sypass43-035 Dumsr in of elbow #14 Yes No No No No No No No line RCS RCS-004 2BCA-14-3" Pressurizer vent bypass43-036 Upstream ofelbow #37 Yes No No No No No No No line '

RCS RCS-004 2BCA-14-3" Pressurizer vent bypass43-037 Downstream ofelbow #37 Yes No No No No No No No linc -

RCS RCS-004 2BCA-14-3" Pressurizer vent bypass43-038 Upstream of elbow #39 Yes No No No No No No No line RCS RCS-004 2BCA-!4-3" Pressurizer vent bypass43-039 Downstream ofcibow #39 Yes No No No No No No No <

line RCS RCS-004 2BCA-14-3" Pressurizer vent bypass43-040 Upstream of rnotor- Yes No No No No No No No line operated nht 2CV-4698-1

Dearadataan Mecharams ,

T-Thermal Fatigue P - Pnmary Wster Stress Carrasson Oncksng (PWSCC) M - Micrainologienny ledluenced Cenomeen ^4fC) F-Flow AccelermandCerromen C-Conomion Cracking 1 - beergranular Streve Cerrosion Cracirs (IGSCC) E- Erossan -Cavitetson 0 - other

'*" FMECA - Degradation Mechanisms C""'"u n Aa Ammw-oi2. Rn. M Page B36 of B36 W eld System 11i Segment Line Number Line Description Namher Weld Imation T C P I M E F O RCS-004 2BCA-14-1" Pressurizer vent line - 43-026 Dowitstream of motor- Yes No No No No No No No RCS 4" section operated vaht 2CV-4730-1 2BCA-14-4" Pressurizer vent line - 43-027 Upstream of motor- Yes No No No No No No No RCS PCS-004 4" sect. ion operated vahr 2CV-4731-2 2BCA-14-4" Pressurizer tent line - 43-030 Downstream of motor- Yes No No No No No No No RCS RCS-004 4" section operated vahr 2CV-4740-2 l

RCS ESS-004 2BCA-14-4" Pressurizer vent line - 43-03I Downstream of 4" x 4" x Yes No No M No No No No 4" section 3" tee #6 (on 4* side)

Pressurizer vent line - 43-032 Upstream of motor- Yes No No No No No No No RCS RCS-004 2BCA-14-4" 4" section operated wahr 2CV-4741-1 h Mechanisms M - Microbiologrally I.dtmenced Cemmon (MIC) F-Flow Accelerated Cer asna T-ThmnalF tigue P - Pnmary Water S:ress Common Cracking (PWsCC)

I-Intergranular Stress Corrouxe Cracking 00 SCC) E-Eranon-Cavitation 0 -Other C-Common Cracking O O O

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t 1 Calculation No. A PENG CALC 012, Rev 00 l- Page C1 of C6 1

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4 l APPENDIX C

'FMECA SEGMENTRISK RANKING REPORT" (Attachment Pages C1 - C6)

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ABB Combustion Engineering Nuclear Operations I

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"-*r" FMECA - Segment Risk Ranking Report C** * ""UC"C*n d""

r, a .r ce 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 RCS-001 226 2BCA-0-6",42-002, 42-003 //43- RCS-N NONE HIGH CAT 4 MEDIUM 2BCA-14-6", 021//06-001,06-2CCA-1-42*, 002,06-005,06-008, 2CCA-10-30",06-009,06-010,06-2CCA-13-3", . 011 // 15-001, 15-2CCA-14-3", 002,15-005, 15-008, 2CCA-15-3",15-009, 15-010 // 27-2CCA-15-4", 001,27-002,27-003, 2CCA-16-2*,27-004,27-005,27-2CCA-2-42", 006,27-007,27-040, 2CCA-29-2",27-041,27-042//26-2CCA-3-30", 001,26-00I A,26-2CCA-30-2", 003,26-004,26-005, 2CCA-31-2",26-006,26-007,26-2CCA-32-2", 036,26-037,26-038 2CCA-4-30*, //26-010,26-011,26-2CCA-5-30", 012,26-013,26-014, 2CCA-6-30", 26",,I5,26-016,26-2CCA-7-30", 017,26-018,26-019, 2CCA-8-30",26-020, 26-020A, 26-2CCA-9-30" 039,26-040,26-041,26-042,26-047,26-048,26-049,26-050,26-05I,26-052,27-010,27-011,27-012,27-013,27-014 27-043,27-044,27-047,27-048,27-049,27-050,27-051,27-052,27-053,27-054,27-O O O

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'N" FMECA - Segment Risk Ranking Report c*d- C-'2 8-*

re a 4ce Degrafstion Number Lines in Welds in Degradation Degradation Mechanism Ceasequence Risk Risk Segment ID of Welds Segment Segment Mechseisms Group ID Category Category Category Rank 055,27-056,27-057,27-058,27-059,27-060,27-061,27-062,27-063,27-064,27-065,27-066,28-041,28-042,28-043,28-044,28-045,28-046,28-047,28-048,28-049,28-050,28-051

//26-043,26-044,26-045,26-046,27-015,27-016,27-017,27-018,27-019,27-045,27-046,28-040//29-002,29-003,29-004,29-005,29-006,29-007,29-008,29-009,29-010,29-011,29-012,29-013,29 4 14,29-015,29-016,29 017,29-018,29-021,29-022,29-023,29-024,29-025,29-042, 29 443,29-044,29-045,29-046,29-047,29-048,29-049,29-050,29 4 51,29-052,29-053 //07-001,07-002,07-005,07-008,07-009,07-010 //36-

8N" FMECA - Segment Risk Ranking Report Na"a""* d">"C8'2 ^""

Page to of C6 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 00I,36-002,36-003,36-004 //10 001,10-002,10-005,10 4 08,10-011, 10-014,10-015,10-016//38-001,38402,38-003, 38-004//39-00I,39-002,39-003,39-004

//35-002,35-003,35-004,35-005,35-006,35-007,35-008,35-009,35-010,35-011,35-012,35-015//Il-00I, Il-002, Il-005, Il-008, Il-009, Il-010,Il-011 //08-001,08-002,08-005,08-008,08-011,08-014,084 15,08-016

//09-001,09-002,09-005,09-008,09-009,09-010, 09-011, 09-012//12-001,12-002, 12 005,12-008, 12-011,12-014, 12-015, 12-016//13-001,13-002, 13-005, ,13-008, 13-009,13-010, 13-01I // 14-001,14-002, 14-005,

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"*" FMECA - Segment Risk Ranking Report c^ w -on e~

raa. cs or cs

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Degradation Number Lines in Welds in Degradation Degradation Mechanism Ceemt Risk Risk Segment ID of Welds Segment Segment Mechanisms Grwup ID Category Category Category Rank-14-008,14-011, 14 014,14-015, 14-016 RCS-002 45 2BCA-t-12",16-001,16 002,16- T RCS-T SMALL HIGH CAT 2 HIGH 2BCA-14-4", 003,16-004,16-005, LEAK 2BCA-14-6",16-006,16 4 07,16-2CCA-15-4", 008,16409,16-010, 2CCA-16-2"16-011, 16-012,16-013,16-014 //43-025,43-029//42-001,43-019,43-020,43-022,43-023,43-024,43-028//28-008,28-009,28-010,28-011,28412,28-013,28-014,28-015, 28416,28-017,28-0I8,28-019,28-020,28-021,28-022,28-023,28-024,28-038, 28-039//29-054,29-055,29-056 l .

'*" FMECA - Segment Risk Ranking Report c h "* d N -o' N "

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Degradation Number Lines in Welds in Degradation Degradation Mechanism Consequence Risk Risk Sepnent ID of Welds Segment Segment Mechanisms Groep ID Category Category Category Rank RCS-003 , 22 2CCA-16-2",29-019,29-020//36- RCS-N NONE MEDIUM CAT 6 LOW 2CCA-29-2", 005,36-006,36-007, 2CCA-30-2",36-008,36-009,36-2CCA-31-2", 010//38-005,38-2CCA-32-2* 006,38-007,38-008,38-009,38-010//39-005,39-006,39 407,39-008,39-009,39-010 //35-013.35-014 RCS404 13 2BCA-14-3", 43 4 33,43-034,43- T RCS-T SMALL MEDIUM CAT 5 hEDIUM 2BCA-14-4" 035,43-036,43-037, LEAK 43438,43-039,43-040 //43-026,43-027,43-030,43-03I,43-032 O O O

4 Calculation No. A-PENG CALC 012, Rev. 00

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' APPENDIX D i

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i QUALITY ASSURANCE VERIFICA TION FORMS I

i ABB Combustion Engineering Nuclear Operations

C:Icul: tion No. A PENG-cal.C-012, R:v. 00 Page D2 of D6 Verification Plan g

Title:

Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the RCS at ANO 2 Document Number: A-PENG-CALC-012 Revision Number: 00 laitner.119ns: Describe the method (s) of verification to be employed, i.e., Design Review, Attemate Analysis, Qualification Testing, a combination of these or an attemative. The Design Analysis Verification Checklist is to be used for all Design Analyses. Other elements to consider in famiulating 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 QP 3.4 of the Quality Procedures Manual).

2. Verification that the appropriate methodology is selected and correctly implemented

3. Verify all design input (as applicable) is appropriately and correctly obtained from traceable sources.

4. Review that the assumptions, results, conclusions, report format, ... etc. are made in accordance with Design Analysis Verification checklist.

A Ver fication PI n prepared by: Approved by:_ qg

D ' vf 0V l Y #

/ eh. ) U Independent Reviewer printed narne and signature 1 Management approver prwelknarne and signature l ABB Combustion Engineering Nuclear Operations

~- - -- . - - - . .- .

C:Icubtion No. A PENG CALC 012 Riv,00 Page D3 of D6 Other Design Document Checklist (Page1of4) inst ructiant? The independent Reviewer is to complete this checkhst 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 relevant for a particular"Other Design Document'. If they are applicable, then the relevant secti n f the Design Analysis Verification Checklist shall be completed and attached to this checklist.

(Sections of the Design Analysis hification Checklist which are not used may be left blank.)

Title:

Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the RCS at ANO-2 Document Numb < r: Revision Number:

A-PENG-CALC-012 '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? 8 2 Are all limitations on the results/ conclusions documented? 8 Documentation Requirements 1, is the documentation legible, reproducible and in a form suitable for filing and retrieving as a Quality @

Record?

11. Is the document identified by title, document number and date? @

lit. Are all pages identified with the document number including revision number? @

IV. Do all pages have a unique page number? @

V. Does the content clearly identify, as applicable:

A. objective Q O B. design inputs (in accordance with QP 3.2) S O C. conclusions E O VI. Is the veriGcation status of the document indicated? @

Vil, if an Independent Reviewer is the supervisor or Project Manager, has the appropriate approval been 29 O documented?

Assumptions

1. Are all assumption identiGed, justified and documented? O O II. Arc all assumptions that must be cleared listed? O 8 A. Is a process in place which assures that those which are CENO responsibility will be cleared? O O O

(,/

B. Is a process in place which assures that those which are the customer's responsibility to clear will be indicated on transmittals to the customer?

O 8 ABB Combustion Engineering Nuclear Operations

C:Icul: tion No. A P:NG CALC 012, R:v. 00 Page D4 of D6 Other Design Document Checkilst (Page 2 of 4) g Assessment of Significant Design Changes Yes N/A

1. Have significant design-related changes that might impact this documerit been considered? O

11. If any such changes have been identified, have they been adequately addressed? O g Selection of Design Inputs

1. Are the design inputs documented? g II. Are the design inputs correctly selected and traceable to their source? g 111. Are references as direct as possible to the original source or documents containing collectioWtabulations of g inputs?

GV. Is the reference notation appropriately specific to the information utilized?

O V. Are the bases for selection of all design inputs documented? O VI. Is the verification status of design inputs transmitted from customers approprintfand documented?

5 O Vll. Is the verification status of design inputs transmitted from ABB CENS appropriate and documented? O S Vill. Is the use of customer-controlled sources such as Tech Specs, UFSARs, etc. authorized, and does the O @

authorization specify amendment level, revision number, etc.?

References i

1. Are all references listed? g

11. Do the reference citations include sufficient information to assure retaevability and unambiguous location g of the referenced material? '

Section 2: Other Potentially Applicable Topic Areas -use appropriate sections of the Design Analysis Verification Checklist (QP 3 A, Exhibit 3.4 5) and attach.

Yes N/A

1. Use of Computer Software g O

2. Applicable Codes and Standards O N

3. Literature Scarches and Background Data O E

4. Methods O E

5. Hand Calculations O E

6. List of Computer Software O E

7. List of Microfiche O O

8. List of optical disks (CD-ROM)

O E

9. List of computer disks

, , ,ff O O -

4: Is.. e k c. dn o j, r o e A bert *vs Ish -

ABB Combustion Engineering Nuclear Operations

Calcul:ti:n No. A.P WG CALC-012 R:v. 00 Page 05 of D6 ID V Other Design Document Checklist .

(Page 3 of 4)

Independent Reviewer's Comments Comment Reviewer's Comment Response Author's Response Response Number Required? Accepted?

I Typo's on page 3,12.16. .i2,33 Yes Typo's have been corrected.

2 On page 8, Figure i previously Yes ne drain line for RCS cold Yes leg 2P32A is connected to showed only %3 cold leg drain lines.

is there is a 4 the letdown line which is addressed as pan of the CVCS.

3 On page 11,the reference to RCPs Yes Concur. Yes being unavailable is not relevant.

Delete.

4 On page 12 (4.1.3b)you refer to Yes Concur. De correct Yes pipe failure occurringduring a wording is" requiring". He demand. Do you really mean '

change has been made

reouiring a demand? accordingly.

5 On page 12 (4.1.4), are the T.S. Yes The T.S. evaluation of Yes requirement and the evaluation pressurizer spray nozzle really the same thing? fatigue is performed to p satisfy the TS requirement

, V (i.e., they are the same thing). Editorial change has been made to reflect this.

6 Page 14. Description of RCS-C-01. Yes Concur Yes Delete "to the SG". It is confusing.

7 Page 30. Are the LTOP line Yes Yes, nese are the line Yes numbers correct? numbers noted on the ISOs.

8 Page 38. Are the indicated Risk Yes Yes. This nomenclature for Yes Segment ID numbers correct? Some the Risk Segment ID was seem truncated. developed and provided as input from ANO2.

9 Page 40. Table claims to contain 23 Yes Concur. Addition has been Yes elements. I see only 22. included.

10 Page 40. Table refers to Figures 7.1- Yes Figure Nos. 7.1-2 & 7.1-1 Yes

, 2 and 7.1 1. Where are they? are not included as pan of this report. Hence, the reference source was included.

lV ABB Combustion Engineering Nuclear Operations

C:Icul: tion No. A PENG CALC-012, Riv, 00 Page D6 of D6 Other Design Document Checklist (Page 4 of 4) g Independent Reviewer's Comments Comment Reviewer's Comment Response Author's Response Respoase Number Required? Accepted?

11 Page A3, etc. Should you fillin the Yes Currently," Consequence Yes space labeled" Cons. Rank"? Rank"is not an input field in the ISIS Database, it is also not calculated. The developers of the database may have included it in the Consequence Information Report form as a place holder for future development.

12 Page Al2, etc. Please explain to me Yes For consistency, a large Yes how we determine that the break break size is selected for size is "large". consequence segments. In the consequence evaluation, there is not basis for the selection because the break size is determined by the degradation mechanism (s).

13 Page A16 refers to 2BCA-0-6'. Is Yes The PSVs are welded to the Yes this correct? The"-0 "looks vessel. There are no line unusual. designators shown on the drawings. it was therefore agreed to use the line number designator as shown.

14 Page A21 refers to 2P32B. Is there Yes There is a drain line Yes also a 2P32A7 associated with cold leg 2P32A, but it is connected to the letdown line. The drain line for cold leg 2P32A is treated as part of the CVCS.

Checklist completed by:

3 Revi$er foBMT G. OAQuI nf nmwo w f[h synmn: &

8/8h7 t,

ABB Combustion Engineering Nuclear Operations

_ _ ____ ______ _-_ ___-__ _______ __________ _ ______ ____________

ML20217E929

Text

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3.1 System Description

9.0 REFERENCES

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Dearadataan Mecharams ,

Title:

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