RS-10-118, Response to Request for Additional Information Related to Relief Requests I3R-01, I3R-02, I3R-03, I3R-04, and I3R-05

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Response to Request for Additional Information Related to Relief Requests I3R-01, I3R-02, I3R-03, I3R-04, and I3R-05
ML101960011
Person / Time
Site: Clinton Constellation icon.png
Issue date: 07/14/2010
From: Hansen J
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-10-118, TAC ME2987, TAC ME2988, TAC ME2989, TAC ME2990, TAC ME2991
Download: ML101960011 (25)
v • d • e


Text

RS-10-1 18 10 CPR 50.90 July 14, 2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Units 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

Response to Request for Additional Information Related to Relief Requests 13R-01, 13R-02,13R-03,13R-04, and 13R-05 (TAC Nos. ME2987, ME2988, ME2989, ME2990, and ME2991)

References:

1. Letter from Jeffrey L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, " Relief Requests Associated with the Third Inservice Inspection Interval," dated December 30, 2009
2. Letter from U. S. NRC to M. J. Pacilio (Exelon Generation Company, LLC), "Request for Additional Information Related to Relief Requests 13R-01, 13R-02,13R-03,13R-04, and 13R-05 (TAC Nos. ME2987, ME2988, ME2989, ME2990, and ME2991)," dated June 14, 2010 In Reference 1, Exelon Generation Company, LLC (EGC) requested approval of relief requests associated with the upcoming third 10-year Inservice Inspection (ISI) Program Interval at Clinton Power Station, Unit 1 (CPS).

During its review of Reference 1, the NRC found that additional information was required to support its review as discussed in Reference 2. The requested information is provided in the attachments to this letter.

July 14, 2010 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter.

Should you have any questions concerning this letter, or require additional information, please contact Mitchel A. Mathews at (630) 657-2819.

Res6}fully,

^ffroy . tansen any er - Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments:

1. Response to NRC Request for Additional Information
2. Clinton Power Station (CPS) Third Inservice Inspection (ISI) Program Interval Risk-Informed (RI)-ISI Tables

Attachment 1 Response to NRC Request for Additional Information Page 1 of 15 fRepuests Associated Withl 13R-01

1. Do the following CPS augmented inspection programs remain unaffected by the Rl-ISI program for the third 10-year inservice inspection interval?
  • High- Energy Line Breaks (USNRC Branch Technical Position MEB 3-1)

EGC Response For the Clinton Power Station (CPS) Third Inservice Inspection (ISI) Interval, the augmented Service Water and Flow-Accelerated Corrosion (FAC) programs remain unaffected by the Risk Informed (RI)-lSI program. The augmented High Energy Line Break (HELB) Break Exclusion Region (BER) program was integrated in the Second Interval RI-ISI program and will remain integrated within the RI-ISI program for the Third ISI Interval. This integration occurred in March 2005 under Revision 1 of the RI-ISI program in accordance with Electric Power Research Institute (EPRI) Technical Report (TR)-1006937, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER)

Programs," Rev. 0-A, including the associated NRC Safety Evaluation.

2. Are the inspection locations in the RI-ISI program that have been developed for the third 10-year interval the same locations as those in the program approved in the NRC staffs April 8, 2002, safety evaluation? If not, please summarize the changes to the program and what caused those changes.

EGC Response The RI-ISl program is required to be and has been maintained as a living program assessing component and configuration changes and major Probabilistic Risk Assessment (PRA) model revisions throughout the Second ISI Interval. As part of the Third Interval update process, the consequence and degradation assignments and resultant component risk rankings have been confirmed or updated, element selections have been adjusted, and the risk impact assessment has been revised. The final RI-ISI evaluation for the previous Second ISI Interval was Revision 2 dated December 2007. The latest evaluation, Revision 3 dated May 2010, is the current evaluation developed as part of the new Third Interval RI-ISI program. The changes in inspection locations from the initial Second Interval RI-ISI program (Revision 0 dated September 2001) to the new Third Interval RI-ISI program are summarized in the table below.

Attachment 1 Response to NRC Request for Additional Information Page 2 of 15 Table 1: CPS Unit 1 Selection Summa

-RiS nn dam . ' Item A fe tincg Chi a

(111^

14% 106 v 0)11e` R 4$J:, R+e . 3)

High 33 37 n Limited Exam Coverage n Plant/Component Modifications n PRA Model Revisions' n Integration of Augmented Break Exclusion Region Medium 41 49 n Limited Exam Coverage n Plant/Component Modifications n PRA Model Revisions' n Revised Makeup Calc n Integration of Augmented Break Exclusion Region Total 74 86 Latest incorporated revision is PRA Model 2006C Limited Exam Coverage - The welds selected for examination were changed in some cases to optimize examination code coverage.

Plant Modifications - As discussed above, the RI-ISI program has been maintained throughout the Second ISI Interval as a living program. Various minor plant modifications were installed throughout the interval and were evaluated for impact to the RI - ISI program, and when applicable, changes to the RI -ISI scope and element selections were made. No major component replacements or new system installations were made during this period affecting the RI-ISI program.

PRA Model Revisions - The CPS PRA Model applicable to the RI-ISI update was revised in March 2007 and issued as Model 2006C. As the model is updated throughout the interval, impact on the RI-ISI program is assessed and the program is updated as necessary.

Revised Makeup Calculation - As part of the Third Interval ISI program, the version of ASME Section XI was updated to the 2004 Edition, No Addenda as required by 1 OCFR50.55a. To implement the new code of record, CPS updated the plant normal makeup calculation for application under IWB - 1220(a). As a result of this update, some smaller diameter lines previously exempted were added into the RI-ISI scope.

Integration of Augmented Break Exclusion Region - In Revision 1 of the RI-ISI evaluation, the augmented break exclusion region inspection (BER or HELB) was added to the RI-ISI program scope. Evaluation, ranking, and selections were made in accordance with EPRI TR - 1006937 and the associated NRC safety evaluation.

Attachment 1 Response to NRC Request for Additional Information Page 3 of 15

3. If there are changes in the inspection locations for the CPS third 10-year interval RI--ISI program, please provide information for the third interval program regarding.

examinationslsystemlcomponentsldegradation mechanismslclass, etc. similar to that provided in Tables 1, 2, 3, 4 and 5 of the original submittal of the RI-ISI program for the CPS second 10-year ISI interval dated October 15, 2001.

EGC Response A summary of the changes to the inspection locations between the original RI-ISI program implemented in the Second ISI Interval and the revised program prepared for the Third ISI Interval is contained in the response to NRC Request No. 2 above. Updated tables similar to those provided in the original submittal of the RI-ISI program are included as Attachment 2 to this response.

4. On page 2 of Enclosure 1 to Attachment 1 to the application, a "current industry peer review of the Clinton PRA" is identified as scheduled for the fourth quarter of 2009.

State whether the results of this industry peer review are available. If the results are available, identify any new findings that are not already listed in "Table 1: Impact of Open Significant PRA Peer Review Findings for the Clinton PRA Model." Provide an impact assessment of any new findings.

EGC Response The results of the recent CPS PRA Peer Review are summarized in Table 4 with a disposition of the associated impacts on the CPS Rl-ISI conclusions. These peer review assessments do not change the conclusions of the CPS RI-ISI risk assessment.

The scope of the peer review results summarized here is with respect to the key Supporting Requirements (SRs) defined for the Mitigation System performance Index (MSPI) program in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6. The SRs identified in NEI 99-02 as relevant to MSPI calculations are also applicable for determining the necessary PRA quality for RI-ISl calculations. MSPI addresses mitigating systems for loss of coolant accident initiating events, as well as other initiating events. These same mitigating systems are important to RI-ISI, and the same PRA elements would be applicable.

In addition, the limited manner of PRA involvement in the RI-ISI process is also reflected in the risk-informed license application guidance provided in NRC Regulatory Guide (RG) 1.174, "Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

Revision 1. Section 2.2.6 of RG 1.174 provides the following insight into PRA capability requirements for this type of application:

There are, however, some applications that, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.

Attachment 1 Response to NFiC Request for Additional Information Page 4 of 15 An example is risk-informed inservice inspection (RI-IS!). In this application, risk significance was used as one criterion for selecting pipe segments to be periodically examined for cracking. During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary.

Therefore, the staff review of plant-specific R!-IS1 typically will include only a limited scope review of PRA technical acceptability.

In the RI-ISI program at CPS, the EPRI RI-ISI methodology as discussed in EPRI TR-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure,"

Revision B-A, is used to define alternative inservice inspection requirements. Plant-specific PRA-derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking, element selection and delta risk evaluation steps.

The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three fundamental components of the EPRI methodology.

First, PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RI-ISI inspection as illustrated in Table 2 below. Broad ranges are used to define these bins so that the impact of uncertainty is minimized and only substantial PRA changes would be expected to have an impact on the consequence ranking results.

Table 2: CCDP and CLERP Thresholds for Inclusion of Welds in RI-ISI Program Consequence Results Binning Groups Consequence Category CCDP Range CLERP Range High CCDP > 1 E-4 CLERP > 1 E-5 Medium 1 E-6 < CCDP < 1 E-4 1 E-7 < CLERP < 1 E-5 Low CCDP < 1 E-6 CLERP < 1 E-7 The risk importance of a weld is therefore not tied directly to a specific PRA result. Instead, it depends only on the range in which the PRA result falls. As a consequence, any PRA modeling uncertainties would be mitigated by the wide binning provided in the methodology.

Additionally, conservatism in the binning process (e.g., as would typically be introduced through PRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to result in a larger inspection population.

Secondly, the impacts of particular PRA consequence results are further dampened by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment. The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment

Attachment 1 Response to NRC Request for Additional Information Page 5 of 15 (and ultimately each element) according to the EPRI Risk Matrix. The Risk Matrix, which equally takes both assessments into consideration, is reproduced in Table 3 below.

Table 3: EPRI Risk Matrix CONSEQUENCES OF PIPE RUPTURE POTENTIAL FOR IMPACTS ON CONDITIONAL CORE DAMAGE PROBABILITY PIPE RUPTURE AND LARGE EARLY RELEASE PROBABILITY PER DEGRADATION MECHANISM SCREENING CRITER IA NONE LOW MEDIUM HIGH HIGH LOW FLOW ACCELERATED CORROSION Category 7 MEDIUM LOW LOW OTHER DEGRADATION MECHANISMS Category 7 Category 6 SSI LOW LOW LOW LOW S NO DEGRADATION MECHANISMS Category 7 Category 7 Category 6 Cat ^'

Thirdly, the EPRI RI-ISI methodology uses an absolute risk ranking approach. As such, conservatism in either the consequence assessment or the failure potential assessment will result in a larger inspection population rather than masking other important components. That is, providing more realism into the PRA model (e.g., by meeting higher capability categories) most likely would result in a smaller inspection population.

These three facets of the methodology reduce the importance and influence of PRA on the final list of candidate welds.

The results of the recent CPS PRA peer review are provided in Table 4 with respect to the key SRs applicable to the MSPI program, as outlined in NEI 99-02 Appendix G.

Attachment 1 Response to NRC Request for Additional Information Page 6 of 15 Table 4:

SUMMARY

OF CPS PRA PEER REVIEW RESULTS AND IMPACT ON RI-ISI Peer Review SR IDt'>,t2I Assessment Peer Review Assessment Basis Im ct on RI-ISI IE-A5 Not Met The SR, as clarified by RG 1.200, requires No impact on RI-CSI conclusions.

(IE-A4) that an evaluation of the potential for each system be performed. While some systems The scope of support system initiating events modeled in have been evaluated, as documented in the CPS PRA includes the typical set of systems as Section 2, there is no evidence that each found in most BWR PRAs. The CPS PRA explicitly system at CPS has been evaluated. Also, includes sixteen (16) support system initiating events.

the basis for excluding loss of electrical area ventilation is not supported with room However, the number and types of support system heat-up calculations. initiators in the PRA is not a critical aspect of RI -ISI risk calculations. In RI-ISI risk calculations a single representative initiator is used in the quantification for each piping segment analyzed and all others are sent to FALSE.

IE-A6 Not Met No systematic evaluation of multiple No impact on RI-ISI conclusions.

(IE-A4) equipment failures, including common cause and routine system alignments, was Refer to response above for SR IE -A5 (i.e., both IE-A5 evident in the documentation. and IE-A6 relate to the processes used to identify support system initiators).

AS-B3 Not Met Nearly all event trees model success of No impact on RI-ISI conclusions.

(AS-B3) emergency core cooling system (ECCS) pump operation following a loss of The bases for ECCS pump failure/success following suppression pool cooling if containment successful containment venting are indeed documented venting is successful. The basis for in the CPS PRA, in sub-section "Successful RPV operation of ECCS pumps is given as Injection and Successful Vent" in the Event Tree Section 6.3.1.1.3 of the USAR which states Notebook and the failure probabilities of the associated that the ECCS pumps are desi g ned to ECCS failure modes (e. Q., basic event 1 SY--

Attachment 1 Response to NRC Request for Additional Information Page 7 of 15 Peer Review SR ldt't,(2) Assessment Peer Review Assessment Basis Im pact on RI-ISI AS-B3 operate at saturation with a suppression STEAMBOUND-, "Vent Causes Steam Binding in ECCS (AS-B3) pool temperature of 212 deg F. However, Suction") are further discussed in Appendix B.13 of the (Continued) containment venting is performed to Component Data Notebook.

maintain containment pressure less than 45 prig which correlates to a saturation ISLOCA and break outside containment (BOO) are temperature of about 250 deg F. No accidents that are important to RI-ISI and play a role in evaluation of the ability of ECCS pumps to determining impact on risk. The operation of ECCS operate at post-containment-venting during these conditions is documented in the Event Tree temperatures was provided. If ECCS pump and Component Data notebooks and is sufficient for RI-operation for such conditions cannot be ISI analyses.

supported, then additional core damage sequences could result. An evaluation of Breaks outside containment are expected to occur in the effects of conditions created by a break containment (which contains little equipment modeled in outside containment on equipment the PRA) or in the Auxiliary Building. (Note; The generic operation is not documented. term Break Outside Containment can also mean break outside the drywell, which is inside the CPS Mark III containment), The Makeup Condensate (MC) and Breaks outside containment and ISLOCA Cycled Condensate (CY) pumps are in a separate events require long-term injection (function building (i.e., the Turbine Building basement). A control XT-CRD) to be successful after an initial rod drive (CRD) pump is normally running, and located period of operation by HPCS or LPCS. in the Turbine Building.

However, the CRD model credits operation of condensate transfer as a suction source The RI-ISI scenario development includes consideration for CRD and condensate transfer could be of impact on equipment surrounding the pipe break failed by the conditions created by the BOC when choosing initiating events and hardware basic or ISLOCA initiators. Also, because the events to represent the scenario. This provides further containment does not have a hard-pipe confirmation that appropriate impacted equipment has vent, failure of the HVAC ducting could been considered.

result in conditions that would cause equipment failures.

Attachment 1 Response to NRC Request for Additional Information Page 8 of 15 Peer Review SR ID(",(" Assessment Peer Review Assessment Basis Im pact on RI-IS1 SY-A22 Not Met No calculations supporting room cooling No impact on RI-ISI conclusions.

(SY-A20) with temporary ventilation are documented.

Better documentation of analysis to The CPS PRA already makes appropriate assumptions demonstrate that rated or design regarding the need for room cooling and explicitly capabilities are not exceeded is needed models room cooling in certain areas. The existing within the system notebooks. modeling provides representation of key room cooling dependencies such that the overall risk characterization is adequate.

Furthermore none of the piping in support of Room Cooling functions is in the scope of the Rl-lSI program.

Attachment 1 Response to NRC Request for Additional Information Page 9 of 15 Peer Review SR ID{'Mat Assessment Peer Review Assessment Basis Im p act on RI-1S1 HR-H3 Not Met Dependent human error probabilities No impact on RI-ISI conclusions.

(HR-H3) (HEPs) have not been evaluated correctly.

Rather than replacing groups of HEPs with This finding appears to be a misunderstanding of the a single assumed value and HEP in the CPS dependent HEP analysis and a reviewer preference post processing recovery file, a for the EPRI CAFTA "RECOVER" function instead of the dependency analysis should be performed "REPLACE" function used in the CPS PRA (either based on BHEP timing. The CPS HRA approach is common and acceptable in the PRA documentation claims that all operator industry). An HEP dependency analysis was performed recovery actions are included directly in the for the CPS PRA (and numerous dependent HEPs are fault trees and event trees and that no explicitly incorporated into the PRA quantification) and is recovery actions were placed in the post documented in the CPS PRA HRA Notebook.

processing recovery file. If so, then the dependent operator recovery HEPs are The CPS PRA HRA does incorporate both methods (i.e.,

incorrectly evaluated. direct incorporation of dependent HEPs directly info fault tree logic, and use of recovery file) as appropriate. The incorporation of the dependent HEPs into the quantification process is discussed in Section 5.3 of the HRA Notebook, which states:

"The treatment of dependencies in the CPS model is performed in a number of ways. The principal ways are:

  • Explicit modeling of dependencies directly in the fault trees
  • Direct modeling of HEPs that have high conditional failure probabilities without the use of QRECOVER
  • Use of the CAFTA Utility, QRECOVER, to substitute a dependent HEP for combinations of HEPs appearing in outsets"

Attachment 1 Response to NRC Request for Additional Information Page 10 of 15 Peer Review SR ID(1),(2) Assessment Peer Review Assessment Basis Impact on RI-ISI DA-B1 Met Data is grouped by mission type and by No impact on RI-ISI conclusions.

(DA-B1) (CC I) system, as documented in Appendix A and C of the Component Data Notebook (CPS- RI-ISI is a "risk ranking" application, and Capability PSA-010, Rev. 2). However, it is hard to Category (CC) I is generally sufficient for RI-ISI.

follow how specific Type Codes are Different postulated approaches to grouping components grouped to match basic event updates in data analysis would not impact the RI-ISI conclusions.

between Tables in App. C and Tables in App. E.

QU-B2 Not Met The truncation study shown in Section 3.1 No impact on RI-ISI conclusions.

(QU-B2) of CPS-PSA-013 shows a 20% change in CDF from a truncation value of 1.0E-11 Many of the calculations are CCDP/CLERP calculations (used for the CPS PRA) to 1.0E-012. The use an initiating event frequency of 1.0 and truncation requirement of this SR is to use a truncation value of 1.00E-10. For initiators like small LOCA. break value that will not eliminate dependencies outside containment or large LOCA this is equivalent to of significant cutsets or accident using the base initiating event frequency and a sequences. Since the definition of truncation of 1.00E-12 or E-13.

"significant" given in the PRA standard is the top 95% of cutsets (or accident The EPRI methodology binning process used to sequences), the "significant" cutsets will determining the High, Medium and Low ranking are change by increasing CDF by 20%. No separated by decades. Quantification results impacts basis for why the truncation value of due to lower truncation levels would have little effect on 1.0E-11 maintains the dependencies of RI-ISI ranking results.

significant cutsets was identified in the documentation. Therefore, this SR is As part of this RAI response, a sensitivity using the considered not met. LERF risk metric was performed. The re-quantification of the LERF risk metric for all CPS RI-ISI cases at a quantification truncation level of 1.OOE-12/yr resulted in no change in the PRA rankings provided in the RI-ISI assessment.

Attachment 1 Response to NRC Request for Additional Information Page 11 of 15 Peer Review SR IDt1>,(2) Assessment Peer Review Assessment Basis Impact on RI-ISI QU-B3 Not Met As described in the PRA standard, No impact on RI-ISI conclusions.

(QU-B3) convergence can be considered sufficient when successive reductions in truncation Refer to response above for SR QU-B2.

value of one decade result in decreasing changes in CDF or LERF, and the final change is less than 5%. The truncation study in Section 3.1.2 of CPS-PSA-013 clearly demonstrates that this convergence has not occurred. For example, in Table 3.1-4 of CPS-PSA-013, there is a 20.5%

difference in CDF between 1 E-1 1 and 1E-12 truncation limits. Similarly, convergence of LERF was not demonstrated with a 29% increase from a truncation of 1 E-11 to 1 E-12.

QU-D4 Met Section 4.5 of CPS-PSA-013 provides a No impact on RI-ISI conclusions.

(QU-D4) (CC I) brief comparison of the CPS Level 1 CDF results with the LaSalle 2003A RI-ISI is a "risk ranking" application, and CC I is model. No other comparisons were generally sufficient for RI-ISI. Documented comparisons performed. Because the standard specifies of the base PRA results with other plants have no direct that a comparison with multiple plants is relationship to RI-ISI calculations.

required and only a brief comparison to one plant was performed, this SR is considered unmet.

Table 4 Notes:

(1) The results of the recent CPS PRA peer review are summarized with respect to the key Supporting Requirements applicable to the MSPI program, as outlined in NEI 99-02 Appendix G.

(2) The Supporting Requirement ID shown in parentheses is the ID from NEI 99-02; the associated ID from the current combined ANS/ASME PRA Standard is shown without parentheses.

Attachment 1 Response to NRC Request for Additional Information Page 12 of 15

[Requests Associated Wlthl 13R-05

1. Please provide component drawings of the Class 2 High Pressure Core Spray, Low Pressure Core Spray, and Residual Heat Removal Pumps which identify the pump casing welds subject to the surface examination requirement of Table IWC-2500-1, Category C-G.

EGC Response As shown in Figure 1 below, the majority of the welds associated with the High Pressure Core Spray, Low Pressure Core Spray, and Residual Heat Removal Pumps are located below the dashed line shown in Figure 1 and encased in concrete. The welds on the accessible portions of the pumps as shown in Figures 2, 3, and 4, will continue to be examined according to the Code.

Attachment 1 Response to NRC Request for Additional Information Page 13 of 15 Inaccessible welds encased in concrete Figure 1: Typical Pump Drawing for the CPS High Pressure Core Spray, Low Pressure Core Spray, and Residual Heat Removal Pumps Showing Accessible and Inaccessible Welds

Attachment 1 Response to NRC Request for Additional Information Page 14 of 15 WE1 LD CROSS REFERENCE PSI T.D. OF61 +

I D14-1 2 D*

3 4 4 DM-?

Figure 2: Drawing of the CPS High Pressure Core Spray Pump Showing Accessible Welds WELD M CROSS REFE RENCE PST I' MFG'S M F -5 0M y^ ^ -2 L

4 014- 6 5 OH-4 Fit DM Figure 3: Drawing of the CPS Low Pressure Core Spray Pump Showing Accessible Welds WELD LO.

CROSS REFERENCE PSI 1.01 MFG'S I.O.

OH-1 0H-5 OH -2 0H-6 OH-4 OH-3 0H-7 Figure 4: Drawing of the CPS Residual Heat Removal Pump Showing Accessible Welds

Attachment 1 Response to NRC Request for Additional Information Page 15 of 15

2. Please identify the start date and currently scheduled end date for the third 10-year ISl interval at CPS.

EGC Response The third 10-year ISI interval at CPS started on July 1, 2010, and is scheduled to end on June 30, 2020.

Attachment 2 Clinton Power Station (CPS) Third Inservice Inspection (ISI) Program Interval Risk-Informed (RI)-ISI Tables Page 1 of 6 Table 1: Clinton Power Station (CPS)

Risk-Informed Inservice Inspection RI-ISI) System List

System Description

Main Steam (PMS)

Reactor Core Isolation Cooling (PRI)

Feedwater (PFW)

Reactor Recirculation (PRR)

Low Pressure Core Spray (PLP)

High Pressure Core Spray (PHP)

Nuclear Boiler (PNB)

Reactor Water Cleanup (PRT)

Residual Heat Removal (PRH)

Scram Discharge Volume (PSD)

Standby Liquid Control (PSC)

Reactor Pressure Vessel (AAI, AAP)

NOTE: This table shows the systems containing ISI Class 1 or 2 Examination Category B-F, B-J, C-F-1, or C-F-2 welds as well as Break Exclusion Region (BER) welds.

Attachment 2 Clinton Power Station (CPS) Third Inservice Inspection (ISI) Program Interval Risk-informed (RI)-ISI Tables Page 2 of 6 Table 2: Failure Potential Assessment Summary for CPS2 Localized Corrosion I Flow Sensitive System IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C x

x X x x

x NOTES:

1. Includes nuclear boiler (PNB).
2. This table shows the assessed failure mechanisms for each system. The RISI program addresses the cumulative impact of all mechanisms that were identified in each system.

TASCS - thermal stratification, cycling and stripping, TT - thermal transients, IGSCC - intergranular stress corrosion cracking, TGSCC - transgranular stress corrosion cracking, ECSCC - external chloride stress corrosion cracking, PWSCC - primary water stress corrosion cracking, MIC - microbiologically influenced corrosion, PIT - pitting, CC - crevice corrosion

, E-C - erosion cavitation, FAC - flow accelerated corrosion

Attachment 2 Clinton Power Station (CPS) Third Inservice Inspection (ISI) Program Interval Risk-Informed (RI)-ISI Tables Page 3 of 6 Table 3: Number of Elements (i.e., welds) by Risk Category for CPS2 TOTAL System All Categories NOTES:

1. Includes nuclear boiler (PNB).
2. This table shows the results of the Risk Categorization for CPS. The risk categories are defined in Figure 3-4 of EPRI TR-112657.

Attachment 2 Clinton Power Station (CPS) Third Inservice Inspection (ISI) Program Interval Risk-Informed (R SI Tables Page 4 of 6 Table 4: Number of Inspections by Risk Category for CPS2, a. 4 High Risk Medium Risk I Low Risk All Risk Category 1 1 Category 2 1 Category 3 Category 5 Categories

( Category 6 or 7 Pre - Pre-I System Pre-RISI RISI Pre - Pre-RISI RISI RISI RISI RISI PHP PLP PMS PSC 0

81 19 326 86 NOTES:

1. Includes nuclear boiler (PNB).
2. This table provides a comparison of the RISI element selection to the previous Second Interval's 1989 ASME Section XI and BER programs (Pre-RISI).
3. This table includes the number of welds previously selected for ASME Section XI and BER (Pre-RISI) that are now addressed by the RI-ISI program, but excludes those selections that now default to the augmented programs for IGSCC and FAC.

Exekn _

Risk Informed Inservice Inspection Evaluation FINAL REPORT REVISION 3 May 2010 Clinton Power Station Unit 1

Clinton RISI Final Report Table 7-8 Impact of RISI' and No Inspections on CDF and LERF Due to Pipe Ruptures for Clinton Systems A CDF A LERF Events/Reactor-Year Ev nts/Reactor-Year System RISI RISI No Acceptance No Acce ptance Inspection RISI Criterion Inspection Criterion 2.52E 1.10E-10

-10 2.52E-10 1.41 E-10 1.41 E-10 1.00E-07 1.07E-1 1.170E-08 "

3.17E-10 3.06E-10 3.68E-10 -1.07E-11 5.16E-11 1.00E-07 -9.05E-1 1,00E-08 1.19E-08 1.08E-08 1,38E-08 -1.08E-09 1.89E-09 1.00E-07 -6,64E-1 1 1 00E-08 7.48E-1 0 7.21 E-10 1.23E-09 -2.64E-11 4.85E-10 1.00E-07 -6.47E-12 1.00E-08 PLP 8.75E-10 7.86E-10 1.72E-09 -8,91 E-11 8.49E-10 1.00E-07 -7,17E-12 1.00E -0 8 PMS 2. 31 E-09 2.79E-09 2.80E-09 4.79E-10 4.93E-10 1.00E-07 7.36E-1 1.00E-08 PNB 4 .32E-11 5,31 E-11 5.31 E-11 9.96E-12 9.96E-12 1.00E-07 7.11 E-1 1.00E-0 PRH 1.07E-08 1.03E-08 1.25E-08 -4.63E-10 1.73E-09 1.00E-07 -1.49E-1 1.00E-08 PRI 3.22E-10 3.80E-10 4.42E-10 5.75E-11 1.19E-10 1.00E-07 7.58E 5.52E-1 1 1.00E-08 PRR 1.54E-08 1,86E-08 1.96E-08 3.18E-09 4.24E-09 1.00E-07 2.41 E-10 3.21E-10 1.00E-08 PRT 1. 00E-08 1. 12E-08 1. 12E-08 1.24E-09 1.24E-09 1.00E-07 5.51 E-10 5.51 E-10 1.00E-08 PSC 3.26E-09 3.06E-09 3.26E-09 -1.94E-10 O.OOE+00 1.00E-07 -5.83E-12 0.OOE+00 1,00E-08 PSD 1.00E-14 1.00E-14 1.00E-14 1.00E-14 1.00E-14 1.00E-07 1,00E-14 1.00E-14 1.00E-08 Total 5.60E-08 5.92E-08 6.72E-08 3.25E-09 1.13E-08 1.00E-06 7.76E-1 0 1.39E-09 1,00E-07

Attachment 2 Clinton Power Station (CPS) Third Inservice Inspection (ISI) Program Interval Risk-Informed (Rl)-ISI Tables Page 5 of 6 Table 5: Impact of RISI and No Inspections on CDF and LERF Due to Pipe Ruptures for CPS Systems System CDF A CDF A LERF Events/Reactor-Year Events/Reactor-Year Events/ Reactor-Year System Pre-RISI I RISI No No Acceptance No Acceptance RIS1 RISI Inspection Inspection Criterion Inspection Criterion AAI 1.10E-10 2.52E-10 2.52E-10 1.41 E-10 1.41E-10 1.00E -07 I 1.07E-11 1.07E-11 1.00E-08 AAP 3.17E-10 3.06E-10 3.68E-10 -1.07E-11 5.16E- 11 1.00E-07 I -9.05E-13 14.37E-12 1 .00 E-08 PFW 1.19E-08 1.08E-08 1.38E-08 -1 .08E-09 1.89E-09 1.00E-07 -6.64E-11 1.82E-10 1.00E-08 PHP 7.48E- 10 7.21 E-10 1.23E-09 -2.64E-1 1 4.85E-10 1.00E-07 -6.47E-12 3.45E-11 1.00E-08 PLP 8.75E-10 7.86E-10 1.72E-09 -8.91 E-1 1 8.49E-10 1.00E-07 -7.17E-12 3.47E-11 1.00E-08 PMS 2.31 E-09 2.79E-09 2.80E-09 4.79E-10 4.93E-10 1.00E-07 7.36E-1 1 8.71 E-11 1.00E-08 PNB 4.32E-11 5.31 E-11 5.31 E-11 9.96E-12 9.96E-12 1.00E-07 7.11 E- 13 7.11E-13 1.00E-08 PRH 1.07E-08 1.03E-08 1.25E-08 -4.63E-10 1.73E-09 1.00E-07 -1.49E-1 1 1.08E-10 1.00E-08 PRI 3.22E-10 3.80E-10 4.42E-10 5.75E-11 1.19E-10 1.00E-07 7.58E-13 5.52E-11 1.00E-08 PRR 1.54E-08 1.86E-08 1.96E-08 3.18E-09 4.24E-09 1.00E-07 2.41 E-10 3.21 E-10 1.00E-08 PRT 1.00E-08 1.12E-08 1.12E-08 1.24E-09 1.24E-09 1.00E-07 5,51E-10 1 5.51E-10 1.00E-08 PSG 3.26E-09 3.06E-09 3.26E-09 -1.94E-10 0.00E+00 1.00E-07 -5.83E-12 1 0.00E+00 1.00E-08 PSD 1.00E-14 1.00E-14 1.00E-14 1.00E-14 1.00E-14 1.00E-07 1.00E-14 I 1.00E-14 1.00E-08 Total 15.60 E-08 5.92E-08 6.72E-08 3.25E-09 1.13E-08 1.00E -06 17.76E-10 1.39E-09 1.00E-07

Attachment 2 Clinton Power Station (CPS) Third Inservice Inspection (ISI) Program Interval Risk - Informed (RI)-ISI Tables Page 6 of 6 Table 6: Impact of RISI and No Inspections on CDF and LERF Due to BER Pipe Ruptures for CPS Systems System CDF A CDF A LERF System No No Acceptance No Pre-RISI RISI RISI Acceptance Inspection Inspection RISI Criterion Inspection Criterion PFW I 9.02E-11 1.23E-10 2.07E-10 3.24E-11 1.17E-10 1.00E-07 1.88E-11 4.79E-11 1.00E-08 PHP 4.58 E-11 1.02E-10 1.05E-10 5.64E-11 5.88E-1 1 1.00E-07 1.61E-13 2.53E-12 1.00E-08 PLP 2.59E-10 2.59E-10 5.91 E-10 1.52E-13 32E-10 1.00E-07 4.55E-14 2.42E-12 1.00E-08 PMS 4.45E-1 1 8.88E-1 1.02E-10 4.44E-1 1 5.78E-1 1 1.00E-07 4.04E-11 5.39E-11 1.00E-08 PRH I 9.12E-12 I 1.62E-11 2.08E-11 7.05E-12 1.17E-11 1.00 E-07 2.03E-12 4.42E-12 1.00E-08

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