ML051790216

Response to Request for Additional Information Regarding Risk-Informed Inservice Inspection Program Plan

ML051790216
Person / Time
Site:	Kewaunee
Issue date:	06/17/2005
From:	Gaffney M Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-05-069, TAC MC2537
Download: ML051790216 (24)

Text

Committed oNucle 9) Operated byKewaunee Operated Nuclear Management Power Plant Nuclear Company, LLC June 17, 2005 NRC-05-069 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Kewaunee Nuclear Power Plant Docket 50-305 License No. DPR-43 Response To Request For Additional Information Regarding Risk-Informed Inservice Inspection Program Plan (TAC No. MC2537)

References:

1) Letter from Thomas Coutu (NMC) to Document Control Desk (NRC),

"In-service Inspection Program for Fourth Inspection Interval", dated December 16, 2003

2) Letter from Carl F. Lyon, (NRC) to Thomas Coutu (NIMC), "Kewaunee Nuclear Power Plant - Request For Additional Information Regarding Risk-informed Inservice Inspection Program Plan (TAC NO. MC2537),"

dated October 14, 2004

3) Letter from Craig W. Lambert (NMC) to Document control Desk (NRC),

"Response To Request For Additional Information Regarding Risk-Informed Inservice Inspection Program Plan (TAC No. MC2537)" dated February 18, 2005 In reference 2, the Nuclear Regulatory Commission (NRC) staff requested additional information concerning the Nuclear Management Company, LLC (NMC) Inservice Inspection Program for Fourth Inspection Interval submittal dated December 16, 2003 (Reference 1). is the NMC response to the NRC request for additional information, which supercedes the response given in reference 3. Enclosure 1 reiterates the NRC staffs request for additional information then adds NMC's response. Based on communications with the NRC staff, the response submitted in reference 3 is hereby resubmitted except for changes to responses to questions 3.5(a), 3.6, 3.10(a), and 3.10(b). Enclosure 2 is Appendix F, "Augmented Examination Program," of the Inservice Inspection Program Plan as updated based on the enclosed responses to the NRC's questions. r\

Document Control Desk Page 2 As stated in reference 1, the ISI program is a working document and changes can be expected to occur during the implementation phase of the program. Accordingly NMC may need to update the program based on changes to the facility and will update the NRC with program revision or relief requests as necessary.

Summary of Commitments This reconfirms the new commitment in the response to RAI 3.4(b) made in reference 2, makes no revisions to any existing commitments, and confirms Kewaunee Nuclear Power Plant did not use Electric Power Research Institute Safety Evaluation of Topical Report TR-1 00693, "Extension of the EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs," in preparation of the Risk Informed Program.

• KNPP commits to review the final approved MRP guidance and NRC limitations on its use. KNPP will either implement these revisions, or provide justification for relief in accordance with 10 CFR 50.55a(g)(5)(iii).

I declare under penalty of perjury that the foregoing is true and correct. Executed on June 17, 2005.

Michael affy~ey/

Site Vice-Presid Kewaunee Nuclear Power Plant Nuclear Management Company, LLC Enclosures (2) cc: Administrator, Region ll, USNRC Project Manager, Kewaunee, USNRC Resident Inspector, Kewaunee, USNRC Electric Division, PSCW

ENCLOSURE I REQUEST FOR ADDITIONAL INFORMATION FOURTH 10-YEAR INTERVAL RISK-INFORMED INSERVICE INSPECTION PROGRAM PLAN KEWAUNEE NUCLEAR POWER PLANT NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-305

1.0 INTRODUCTION

By letter dated December 16, 2003, Nuclear Management Company, LLC (the licensee), proposed an alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to use a Risk-Informed Inservice Inspection (RI-ISI) Program as an alternative to certain requirements listed in Section Xl for the inspection of ASME Code Examination Categories B-F, B-J, C-F-1 and C-F-2 piping at Kewaunee Nuclear Power Plant (KNPP). The alternative is based on the risk-informed process described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A

'Revised Risk Inforned Inservice Inspection Evaluation Procedure".

Inservice inspection of ASME Code Class 1,2, and 3 components is to be performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code, and applicable addenda, as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The regulation at 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if the licensee demonstrates that (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code, Section Xl, Rules for Inservice Inspection (ISI) of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, which was incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The Code of Record for the KNPP fourth 10-yearinterval inservice inspection program is the 1998 Edition of Section Xl of the ASME Boiler and Pressure Vessel Code, with the 2000 Addenda.

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2.0 SCOPE The licensee's RI-ISI Program alternative was submitted in summary form developed under the guidelines described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A "Revised Risk Informed Inservice Inspection Evaluation Procedure" The RI-ISI Programencompasses all Code Class I and 2 piping at KNPP.

The staff has reviewed the licensee's submittal and, based on this review, determined the following information is required to complete the evaluation.

3.0 ADDITIONAL INFORMATION REQUIRED 3.1 KNPP states that the quantitative risk impact assessment was performed using the 'Simplified Risk Quantification Method" described in Section 3.7 of EPRI Topical Report (TR)-1 12657 and that the PBF [pressure boundary failure] is determined by the presence of different degradation mechanisms. KNPP states that best estimate and upper bound failure frequencies assumed in the delta risk assessments were consistent with assumptions made in original EPRI RI-ISI pilot studies at Vermont Yankee (boiling water reactor) and Arkansas Nuclear One, Unit 2 (Combustion Engineering-pressurized water reactor). Provide justification that you meet the guidance in Fig 3-6 of the EPRI TR-1 12657 and don't need to perform the more realistic estimates for Delta CDF [change in core damage frequency] and Delta LERF [change in large early release frequency].

NMC Response to 3.1 Section 3.7.2 and Figure 3-6 of the EPRI TR both specify that a bounding risk analysis can be performed. Then, if the criteria are not met, a more realistic quantitative analysis should be performed. Section 3.7.2 further states that in the EPRI RI-ISI pilot studies and supporting research project, two approaches for a realistic quantitative analy'sis have been used and successfully applied. One is based on a simple approach with reasonable assumptions ("Simplified Risk Quantification Method"), and another more detailed approach requires more data to apply ("Markov Risk Quantification Method").

Therefore, realistic quantitative estimates for Delta CDF and Delta LERF in this case were performed using the "Simplified Risk Quantification Method." The results showed that system-level criteria were met for each system. (Refer to Table 3.6-1 from reference 1.)

3.2 In the risk impact analysis, KNPP assumes that the RI-ISI probability of detection (POD) for service-induced fatigue cracks resulting from thermal stratification, cycling and striping (TASCS) and thermal transients (TT) degradation mechanisms is 0.9. A recent POD study by EPRI, DOE [Department of Energy],

and NRC, documented in EPRI Guideline MRP-82, developed fatigue crack POD curves for carbon and stainless steel piping based on crack detection data assembled as part of the industry's performance demonstration initiative (PDI).

These results show that, for all thicknesses of carbon/low alloy steel piping and austenitic stainless steelpiping with wall thicknesses < 0.4-inches, the best estimate POD = 0.9 is a reasonable assumption. However, for thicker stainless steel pipes, this assumption is too optimistic with respect to detection of fatigue Page 2 of 15

cracks with through-wall depths identified in Section Xl flaw acceptance standards. For stainless steel pipes with section thicknesses greater than or equal to 0.4-inches, the best estimate and 95% confidence lower bound PODs for 10% through-wall fatigue cracks were approximately 0.8 and 0.7, respectively.

Discuss the impact that these recent studies have on the delta risk input results.

Please describe the risk impact considering upper bound failure frequencies and 95% confidence lower bound POD, as applicable.

NMC Response to 3.2 The risk impact analysis was performed both with and without crediting an improvement in POD. Inthe conservative case (no POD improvement) the acceptance criteria for change in risk was met. A summary of this analysis is provided in NMC-01-343, given in reference 1.

3.3 It is noted that Table 5-2 is intended to summarize and compare new RI-ISI with existing ASME examinations, and this table lists relevant degradation mechanisms for elements (examination locations) by plant system, and includes other relevant information. In order to determine if appropriate examination methods are being correctly applied to target specific degradation, further clarification is necessary. Please abreak-out" the planned methods for examination, i.e., show how many volumetric or surface examinations will be applied as a result of the Ri-ISI process, instead of listing these only as UNDE" or confirm that NDE [nondestructive examination] implies volumetric examination (typically ultrasonic testing). What NDE method will be used for Category 4 items selected for inspection?

NMC Response to 3.3 All RI-ISI examinations will be volumetric. Surface examinations would only be used if the External Chloride Stress Corrosion Cracking degradation mechanism was identified, which was not the case at KNPP. All Category 4 item examinations will also be volumetric.

Surface examinations in addition to the required volumetric examination are currently scheduled on Circumferential Butt Welds during the Fourth Ten-Year Interval in excess of the Risk-Informed Inservice Inspection Program requirements. Surface examinations, when preformed, may permit a credit to the percentage of the examination when 100% access for volumetric examination is not achieved. Surface examinations will also ensure weld integrity when preparation is performed on the weld crown of Circumferential Buff Welds, by weld metal removal in preparation for ASME Boiler and Pressure Vessel Code Section Xl Appendix Vill Ultrasonic Examinations.

3.4 In Section 3 of the KNPP submittal, it is stated that a deviation to the EPRI RI-ISI methodology has been implemented in the failure potential assessment for the damage mechanism of TASCS. For clarification, provide confirmation to the following two items pertaining to the assessment of TASCS:

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a) Confirm that the methodology for assessing TASCS in the proposed RI-ISI program is identical to the materials reliability program (MRP) methodology in EPRI TR- [sic] 000701, "Interim Thermal Fatigue Management Guideline (MRP-24)," January 2001.

NMC Response to 3.4(a)

The methodology used for assessing TASCS at KNPP is identical to that in Guideline MRP-24 for the specific lines covered by that document. In addition, the underlying basis of MRP-24 was used for assessing TASCS for other lines. Thus, the methodology for addressing TASCS is fully consistent with Guideline MRP-24.

b) Currently MRP-24 is an interim document that is undergoing review by EPRI and the NRC. Please confirm that once the final MRP document is published, KNPP will incorporate the applicable final guidance (as approved, restricted, or amended by the NRC) into the KNPP RI-ISI program.

NMC Response to 3.4(b)

KNPP commits to review the final approved MRP guidance and NRC limitations on its use. KNPP will either implement these revisions, or provide justification for relief in accordance with 10 CFR 50.55a(g)(5)(iii).

3.5 Several issues are unclear regarding existing "augmented" examination programs at KNPP and how these may be impacted due to the RI-ISI evaluation:

a) With regard to inspections required as a result of IE Bulletin (IEB) 79-17, Pipe Cracks in Stagnant Borated Water Systems at PWR Plants, the licensee argues that no augmented inspections will be performed on stagnant borated lines because, as stated by the licensee, 'a review performed by KNPP and the historical performance of the weldments to-date signify that these systems will not experience stress corrosion cracking."

Similarly, the licensee stated that no augmented inspections will be performed on pipe segments associated with IE Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems, because examinations that KNPP committed to perform have not revealed any indications.

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It is not clear whether inspections performed as a result of IEB 79-17 and IEB 88-08 constitute augmented examinations. Further, the technical justifications provided by the licensee are not adequate for the staff to conclude that the subject piping segments should not be included in the RI-ISI evaluation. If the examinations performed as a result of the subject IEBs are used as a basis to exclude pipe segments from evaluation under the RI-ISI program, then this may constitute a deviation from the process described in the EPRI Topical Report.

Please discuss the basis and requirements for IEB inspections at KNPP, and the relationship between the IEB inspections and the RI-ISI evaluations performed at KNPP. In the discussion include the extent and frequency of examinations performed as a result of commitments made to NRC, and provide a clearjustification for why the subject piping segments are not included in the RI-ISI evaluation.

NMC Response to 3.5(a)

Based upon Section 3.2.1 of EPRI TR 112657 Rev. B-A, augmented plant programs to address IEB 79-17 and IEB 88-08 concerns are replaced by the RI-ISI program.

Previous examinations performed as a result of the subject IEBs were not used as a basis to exclude pipe segments from evaluation under the RI-ISI program. Therefore, no deviation was taken from the process described in the EPRI Topical Report in this instance. Section 3.2.1 of EPRI Topical Report TR 112657 Rev. B-A is located in the Front Section Titled Safety Evaluation Report Related to "Revised Risk-Informed Inservice Inspection Evaluation Procedure" (EPRI TR-112657, Rev. B July 1999) on pages 4 and 5 under Scope of Program.

3.6 In Section 3.1, KNPP states that additional plant information including the existing plant ISI program were used to define the Class 1 and 2 piping system (RI-ISI evaluation) boundaries. The submittal does not describe what additional plant information has been used nor how this information was used to define the system RI-ISI evaluation scope. Since Class I and 2 piping are well defined, how was the existing ISI program used to define the evaluation scope? Did the RI-ISI evaluation scope include Class 2 piping segments that are normally exempt from examination in the existing ISI program according to IWC-1220? If not, please explain why these portions of IWC-1220 exempt piping have not been included in the RI-ISI evaluation.

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NMC Response to 3.6 The "additional plant information" wording in Section 3.1 is standard wording* used in previous submittals. The "additional plant information" wording is used to describe information related to those piping segments not previously in the ISI program scope.

The RI-ISI evaluation scope was defined consistent with the existing plant ISI program.

As has been the case with all follow-on plant applications using the EPRI RI-ISI method, the Class 2 scope that is normally exempt from examination in the existing ISI program according to IWC-1 220, is also considered exempt from the RI-ISI program. The only piping segments added to the RI-ISI program that were not included in the existing plant ISI program were thin-walled (<3/8") piping segments connected to the Refueling Water Storage Tank (RWST). These piping segments were not exempt according to IWC-1220, but were also not required to be examined under the existing plant ISI program.

The best reference for justification of this statement is located in "Appendix R Risk-Informed Inspection Requirements For Piping" to be published in the ASME Boiler and Pressure Vessel Code Section Xl, Division 1 2005 Addenda Approval Date: October 8, 2004. The referencing location is Nonmandatory Appendix R Supplement 2 Risk Informed Selection Process - Method B Section 2.0 (Boundary Identification) on page 26 which states:

The Owner shall define the system boundaries included in the scope of the risk-informed inspection program evaluation. Within each system boundary, the risk-informed evaluation may include Class 1, 2, or 3 piping defined in the deterministic inservice inspection program, if applicable, and piping outside the current deterministic program examination boundaries, if applicable. As a minimum, piping, or portions thereof, included for evaluation shall be based on the deterministic program for Class 1, 2, or 3, examination boundaries, if applicable, and are determined in accordance with the requirements of IWA-1 320, and limited by exemptions of IWB-1 220, IWC-1 220, and IWD-1220. When Examination category C-F-1 or C-F-2 piping is included, the piping exempt from NDE under the requirements of Table IWC-2500-1 due to nominal wall thickness limitations shall be evaluated.

3.7. From the staffs review of the safety evaluation report(SER) of the Kewaunee Nuclear Power Plant (KNPP) individual plant examination (IPE), it appearsthat there were no identified weaknesses with the IPE methodology. Please confirm that this is correct, or otherwise indicate 1) what weaknesses were identified and

2) what was done to correct the identified weaknesses (in subsequent revisions to the KNPP IPE), or why the uncorrectedweaknesses are not relevant to the RI-ISI application.

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NMC Response to 3.7 There were six weaknesses identified in the IPE review. These are as follows:

1. Spray was not considered in internal flooding: This weakness was also identified by the 2002 Peer Review team. As stated in the submittal, flooding was addressed using a separate analysis, not by using the flooding PRA.

2. Justification for not including certain phenomena in the containment event trees is absent: This justification does exist, but was not included in the original submittal. This weakness was resolved by clarifying the Level 2 documentation.

3. The link between plant damage states and containment performance is lacking: This weakness was resolved by clarifying the Level 2 documentation.

4. The definition of a vulnerability is vague: This weakness relates to identifying vulnerabilities rather than determining importance of components. It therefore has no effect on risk-informed ISI.

5. The timing of human interactions (His) was not adequately addressed:

This weakness was also identified by the 2002 Peer Review team. Refer to the answer to question 3.8b.

6. Dependency between His may not be complete: This weakness was addressed subsequent to the SER and prior to the Westinghouse Owners Group (WOG) peer review. Each combination of two or more His within a cutset is now analyzed. The WOG team evaluated this methodology and found it to be appropriate.

3.8 In Section 1.2 of your submittal, you note that an industrypeerreview was completed in June 2002, which concluded that 'the Kewaunee probabilistic risk assessment (PRA) could be effectively used to support applications involving risk significance determinations supported by deterministic analyses once the Facts and Observations (F&Os) noted in the report are addressed." You then provide details for five Category A F&Os. Below are questions pertaining to two of these.

a) aLong-term condensate storage tank inventory is not appropriately modeled for the loss of service water scenario. The resolution of this issue showed that it did not have a major effect on results."

Please explain why the impact on the consequence ranking of pipe segments would not be affected by the resolution of this F&O.

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NMC Response to 3.8(a)

Long-term cooling was addressed in PRA revision 0101.016, completed January 31, 2003, by adding it to the auxiliary feedwater fault tree. For auxiliary feedwater to be successful in a loss of service water event, the condensate storage tanks must be cross-tied to the reactor makeup water storage tanks.

Since this is a relatively simple action and uses passive components, it adds minimal additional risk. As a result of this model change, the CDF for a loss of service water increased by 19%. This is not sufficient to change any consequence ranking.

b) 'The bases for the time windows for human actions are not well defined.

Work is in progress on resolving this F&O. Preliminary results show that the human error probabilities (HEPs) in the model tend not to be greatly affected by the new time windows."

Please indicate whether or not there have been any changes to the above preliminary results. If so, please identify any significant increases in HEP values, and whether or not these increased HEP values may have an impact on the consequence ranking of pipe segments.

NMC Response to 3.8(b) I The following HEPs significantly increased from the value used in the 0101 model.

Human Error ID Description Old HEP New HEP 47-ORT-HE OPERATOR FAILS TO TRIP REACTOR MANUALLY 1.OOOE-04 9.919E-03 02-PM-SWIA1-AE OPERATOR FAILS TO RESTORE SW PUMP Al AFTER TEST 1.OOOE-05 7.991 E-04 02-PM-SWlA2-AE OPERATOR FAILS TO RESTORE SW PUMP A2 AFTER TEST 1.OOOE-05 7.991 E-04 02-PM-SWlBl-AE OPERATOR FAILS TO RESTORE SW PUMP B1 AFTER TEST 1.OOOE-05 7.991 E-04 02-PM-SWI B2-AE OPERATOR FAILS TO RESTORE SW PUMP B2 AFTER TEST 1.OOOE-05 7.991 E-04 05ALO-FWIOAB-HE OPERATOR FAILS TO LOCALLY OPEN FW BYPASS VALVE 8.240E-03 4.340E-01 05B-AFW-SLB-HE OPERATOR FAILS TO RESTART AFW PUMPS IN STEAM LINE BREAK 4.434E-04 1.925E-02 36-OBF-HE OPERATOR FAILS TO ESTABLISH BLEED ANDFEED 6.684E-04 2.451E-02 34-ISL-DIAG-HE OPERATOR FAILS TO DIAGNOSE ISL 3.OOOE-04 6.841E.03 36-SGTRDIAG-HE OPERATOR FAILS TO DIAGNOSE SGTR 1.000E-04 1.123E-03 33-2TRN-REC-HE OPERATOR FAILS TO ESTABLISH RECIRC (1 OF 2 TRAINS) 2.048E-03 2.133E-02 None of these HEPs changes the consequence ranking of any consequence in the submittal.

In order to examine all of the model changes that resulted from peer review F&Os, the consequence analysis was reanalyzed using the current model (0403A, Released October, 2004). This model reflects all the new HEPs and addresses all the F&Os in the model as defined in the table given in the response to question 3.9(a).

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The Small LOCA consequence moved from High to Medium in the current model, but this change is conservatively neglected at this time. Small LOCA is still treated as a High consequence.

3.9 Your submittal states that, from the June 2002 Westinghouse Owners Group Peer Review, there were 49 Category B F&Os.

a) Please provide these F&Os, and identify those which have already been addressed in the 0101 PRA model.

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NMC Response to 3.9(a)

Since the 0101 model was released before the peer review, none of the F&Os had been addressed in this model. In the current model, the 0403A model, 4 of the 5 Category A F&Os and 41 of the 49 Category B F&Os have been addressed. The following table provides the status of each of these F&Os.

Peer Review Facts and Observations F&O Category Description In Model? How Addressed No.

IE-1 B Loss of vent initiating event No ISI applies to piping, not ventilation.

not well discussed.

IE-2 B ransient PORV opening Yes Only small effect on LOCA frequency.

does not include HI.

IE-3 B ISLOCA does not use latest Yes Changes resulted in decrease of NUREG. ISLOCA frequency.

IE-9 B Steam line break uses worst Yes This is conservative and SLB is still a case assumptions. small contributor.

IE-1 1 B Need to look at variance term Yes Changes resulted in decrease of in ISLOCA IE frequency. ISLOCA frequency.

IE-12 B Look at repair time in initiator Yes Resulted in slight decrease of loss of fault trees. service water frequency.

IE-14 B Spurious SI and effect on Yes Small effect because the pumps still MFW not modeled. can be started manually.

AS-1 A CST inventory for AFW not Yes Resulted in very minor change in modeled for loss of SW. CDF.

AS-2 B Failure to trip RXCP not Yes This was resolved with a calculation.

modeled as a seal failure. No model change was needed.

AS-3 A Time phasing and type of Yes Change in SBO CDF was very slight.

LOSP not modeled for SBO.

S-4 B Establish safe stable end Yes Removal of RWST refill model from states for RWST Refill and ISLOCA resulted in only slight ISLOCA. increase.

S-5 B Look at other initiators Yes TWS is a very minor contributor even leading to ATWS. with this change.

AS-8 B References for WOG 2000 Yes Documentation issue only.

model incorrect.

TH-1 B Update Appendix 3B based Yes Documentation issue only.

on new MAAP runs.

TH-2 B Insufficient guidance for Yes Documentation issue only.

success criteria determination.

TH-4 B HVAC calcs need to be No See response to question 3.9 (b).

redone.

TH-5 A There needs to be a better Yes See response to question 3.8(b).

basis for time of Hls.

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Peer Review Facts and Observations F&O Category Description In Model? How Addressed No.

TH-6 B Mission times for HPI, HPR, Yes The mission times used were LPR need to be more event- conservative, the results changed only specific. slightly.

TH-8 B MAAP limitations need to be Yes Documentation issue only.

examined in more detail.

TH-1 1 B MAAP is not properly Yes There is no evidence of code controlled. problems.

SY-1 B CST inventory for AFW not Yes Resulted in only a slight change in modeled. transient CDF.

SY-3 B CCW for RHR pumps not Yes This was resolved with a calculation.

modeled if on recirculation No model change was needed.

SY-5 B Biofouling of SW and steam Yes Documentation issue only.

binding of AFW pumps not modeled.

SY-8 B CST makeup to condenser Yes It is a passive system; it has low not modeled. failure rates and will minimally change the CDF.

SY-9 B Miscellaneous systems No Documentation issue only.

notebook needs more detail.

SY-1 0 B Normally open valves need No It is a minor contributor but needs to to be modeled. be addressed for completeness.

DA-2 B Some common cause basic Yes They are not different by a large events are overly amount. The CDF change was very conservative. slight.

DA-7 B Reason for variable time Yes Documentation issue only.

periods for T&M needs to be documented.

DA-10 B LOSP due to grid instability Yes It is modeled as part of LOSP following a trip not modeled. frequency. Documentation issue only.

HR-1 B Operators need to be more Yes See response to question 3.8(b).

involved in HRA.

HR-2 B No screening for Yes Miscalibration errors were modeled miscalibration errors, no and had only a slight effect.

guidance for post initiator screening.

HR-5 B Wrong branch used for 36- Yes Diagnosis failure results in core SGTR-DIAG-HE. damage, so piping importance is irrelevant.

HR-6 B Accumulator refill modeled Yes Not a piping issue. An HRA issue only.

too conservatively.

HR-7 B HEP time windows not clear. Yes See response to question 3.8(b).

DE-2 B Loss of letdown not modeled Yes Affects CVCS and CCW only. CCW for loss of CCW and CVCS (other than as an ISLOCA

__initiator) not subject to ISI.

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Peer Review Facts and Observations F&O Category Description In Model? How Addressed No.

DE-3 B New flooding walkdown Yes Flooding was reviewed in the ISI needed. submittal.

DE-6 B There are no structural Yes hese calculations have been calculations for flooding reviewed and are referenced in our doors. analysis.

DE-7 A Numerous errors in the No Flooding was reviewed in the ISI flooding analysis. l submittal.

DE-8 B HRA in flooding not No Flooding was reviewed in the ISI considered. submittal.

DE-9 B No operating experience No Flooding was reviewed in the ISI review was performed for submittal.

flooding.

DE-10 -B Service water frazil ice failure Yes It really is modeled. Documentation not adequately discussed. issue only.

DE-1 I B Spray not adequately No Flooding was reviewed in the ISI addressed in flooding. submittal.

ST-1 B Ability of valves used for Yes This is now addressed. The difference ISLOCA isolation to hold is very slight.

RCS pressure not addressed.

QU-1 A Accumulator refill modeled Yes Not a piping issue. An HRA issue only.

too conservatively.

QU-2 B Guidance documents are Yes Documentation issue only.

needed.

QU-3 B There needs to be a better Yes Documentation issue only.

description of the results.

QU-5 B Logic loops need to be better Yes Documentation issue only.

documented.

QU-7 B More sensitivities are Yes Documentation issue only.

needed.

QU-1 0 B No parametric uncertainty Yes Documentation issue only.

analysis performed.

L2-2 B EAL bases needs to be Yes The sequences concerned did not reviewed for MAAP runs. involve piping failures.

L2-6 B Level 2/LERF quantification Yes Documentation issue only.

is not well documented.

MU-3 B Software control is weak. Yes Documentation issue only.

MU-4 B Permanent Applications form Yes Documentation issue only.

needs to be completed.

MU-6 B No discussion of what review Yes Documentation issue only.

of notebooks means.

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b) For those F&Os not already addressed in this model, you indicate that none of them would result in a significant increase in failure rates or consequences. Please provide your rationale that led to this conclusion for each of these F&Os.

NMC Response to 3.9(b)

In order to determine the effects of addressing the F&Os, the current model was examined. Most consequence rankings remain the same, and one moves to a lower category (High to Medium). The effects of this change are discussed in the answer to question 3.8(b).

Of the one Category A and eight Category B F&Os not addressed in the current model, the one Category A and four of the eight Category B events deal with flooding. Since flooding was resolved separately in the consequence analysis and the PRA was not relied upon, these F&Os have no effect on risk-informed ISI.

One of the remaining four F&Os deals with the level of detail in the Miscellaneous Systems Notebook. Since this is a documentation issue only, it has no effect on risk-informed ISI.

One of the remaining F&Os deals with the lack of passive failures, such as locked open motor operated valves transferring closed in the model. Although this F&O will be addressed in a future model, it is not important to the base model. The purpose of modeling these valves is to provide flexibility for the online risk assessment model, so any segment of piping can be removed from service.

The final two remaining F&Os deal with heating ventilation and air conditioning (HVAC) system modeling. The calculations used to credit HVAC in the model and to screen out HVAC initiators are lacking or are overly conservative. Since HVAC systems involve ducting rather than piping and the service water failure consequence is conservatively used for the service water lines to HVAC systems, these systems are not important to risk informed ISI.

3.10 Section 2.2 of your submittal lists one augmented inspection program (Generic Letter 89-08) that was considered during the development of this RI-ISI application. You also provided with your submittal a more detailed description of augmented inspection programs at KNPP in Appendix F of Structural Integrity Calculation/File No. NMC-01-343, "Risk Impact Analysis for the Kewaunee Nuclear Power Plant'; Revision 0, July 29, 2003. With the exception of your information about the relationship between Generic Letter 89-08 and the RI-ISI program, it appears that the other augmented inspection programs are completely independent of the RI-ISI program.

a) Please confirm that none of the programs listed in the above Appendix F is subsumed into the Rl-ISI program, or otherwise identify which one [sic]

are subsumed.

Page 13of15

NMC Response 3.10(a)

Appendix F of the Kewaunee Nuclear Power Plant Fourth Ten Year Inservice Inspection (1SI) Program 2004-2014 will be revised (Reference Attached) to remove all references to Nuclear Regulatory Commission Information Notice 93-20 which is related to IE Bulletin 79-13, Cracking in Feedwater System Piping, IE Bulletin 79-17, Pipe Cracks in Stagnant Borated Water Systems at PWR plants and IE Bulletin 88-08 Thermal Stresses in Piping Connected to Reactor Coolant Systems. This is due to the fact that these Bulletins were subsumed by the KNPP Risk Informed ISI Program as stated in EPRI Topical Report TR 112657 Rev. B-A Section 3.2.1 located in Section Titled Safety Evaluation Report Related To " Revised Risk-Informed Inservice Inspection Evaluation Procedure" (EPRI TR-112657, Rev. B July 1999) on pages 4 and 5 under Scope of Program. A search of the Kewaunee Nuclear Power Plant commitments to the Nuclear Regulatory Commission revealed that KNPP had made no Commitments to the NRC for Information Notice 93-20, which is related to IE Bulletin 79-13 and IE Bulletin 79-17. A search of the Kewaunee Nuclear Power Plant commitments to the Nuclear Regulatory Commission revealed that KNPP had met all required commitments related to IE Bulletin 88-08.

Additionally:

a. A search of the Program implemented at KNPP in 1979 for IEB-79-13 revealed that there were 2 welds associated with Cracking in Feedwater System Piping.

These 2 welds are ISI Code Class 2 and are integrated into the current KNPP RI-ISI Program.

b. A search of the Program implemented at KNPP in 1980 for IE Bulletin 79-17, Pipe Cracks in Stagnant Borated Water Systems at PWR Plants revealed that all segments listed as Stagnant Borated Water Systems in 1980 are currently part of the ISI Class 1 and Class 2 Systems integrated into the KNPP RI-ISI Program except for the following System identified as ISI Class 3 in 1980:

1. 8" Chemical and Volume Control Piping from the Boric Acid Tanks 1A and 1B through 8" Valves SI-1A and SIlAB to 8" Class 2 Valves SI-2A and Sl-28. The 8" Chemical and Volume Control System from the Boric Acid Tanks through 8" Valves SI-1A and SI-1B to the Class 2 Valves Sl-2A and Sl-2B are currently classified as Non Code Class due to a reclassification of the Boric Acid Tanks under DCR 2786 in 1996 from ISI Code Class 3 to ISI Non Code Class based on identifying the Safety Related source of boron for reactor shutdown as the Refueling Water Storage Tank in lieu of the Boric Acid Tanks.

c. A search of the Program implemented at KNPP in 1989 for IEB-Bulletin 88-08 revealed that there were 3 welds associated with Thermal Stresses in Piping Connected to Reactor Coolant Systems. These 3 welds (2 Residual Heat Removal and 1 Pressurizer Surge) are ISI Code Class 1 and are integrated into the current KNPP-RI-ISI Program.

Page 14 of 15

Kewaunee will also include in Appendix F a reference to Nuclear Regulatory Commission Bulletin 87-01: Thinning of Pipe Walls in Nuclear Power Plants and Nuclear Regulatory Commission Generic Letter 89-08: Erosion/Corrosion-Induced Pipe Wall Thinning b) Please confirm that none of the piping-related nondestructive examination (NDEs) to be performed for the Appendix F augmented inspection programs is being credited toward the count of NDEs (locations) for RI-ISI. Otherwise, please describe which programs you intend to do this with, how many welds from each program you intend to credit toward the count of NDEs for RI-ISI, and why you feel that the prescribed augmented inspection NDE will adequately suffice for the examination method(s) required under RI-ISI.

KNPP Response 3.10 (b)

• This is to confirm that none of the additional examinations, with the exception of the Flow Accelerated Corrosion (FAC) Program per Nuclear Regulatory Commission Generic Letter 89-08 requirements, performed in the Appendix F Augmented Examinations of the Kewaunee Nuclear Power Plant Fourth Ten-Year ISI Program 2004-2014 will be credited to the Risk Informed Program.

Page 15 of 15

ENCLOSURE 2 APPENDIX F AUGMENTED EXAMINATION PROGRAM

'-7 6 pages follow

Appendix F Augmented Examination Programs Introduction Augmented examinations are those examinations that are performed above and beyond the requirements of ASME Boiler and Pressure Vessel Code Section XI, examinations that are governed by Kewaunee Nuclear Power Plant Updated Safety Analysis Report and Technical Specifications or required to be performed to ASME/ANSI OM Standard Part 4 as referenced in ASME Boiler and Pressure Vessel Code Section XM. Below is a summary of those examinations performed by the Kewaunee Nuclear Power Plant that are not specifically addressed by ASME Boiler and Pressure Vessel Code Section XI or the KNPP Risk Informed Program (located in Section 10), that will be performed during the Fourth Ten Year Inspection Interval.

Program Summary Augmented examinations performed at the Kewaunee NuclearPowerPlanton acontinuous or ongoing basis are as follows:*

a. NRC Order's and NRC Bulletins

i. Nuclear Regulatory Commission Interim Head Inspection Requirements NRC Order EA-03-009 ii. Nuclear Regulatory Commission NRC Bulletin 2003-02: Leakage From Reactor Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity iii. Nuclear Regulatory Commission NRC Bulletin 87-01: Thinning of Pipe Walls in Nuclear Power Plants

b. Generic Letters

i. Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Coolant Boundary Components in PWR Plants." Reference WPSC letter, NRC-88-077, dated July 3, 1989.

ii. Generic Letter 89-08, Erosion/Corrosion - Induced Pipe Wall Thinning

c. KNPP Updated Safety Analysis Report

i. Section 4.2.2, Reactor Coolant Pump Flywheels

d. KNPP Plant Technical Specifications

i. TS 4.2.b, Steam Generator Tubes ii. TS 4.14, Testing and Surveillance of Shock Suppressors (Snubbers)

F-1 Rev. 1

Appendix F Augmented Examination Programs

e. Kewaunee Nuclear Power Plant Engineering Programs Inservice Inspection

i. Alloy 600/82/SA182 Weld Program ii. Steam Generator Feedwater Nozzle to Pipe Welds

2. Program Implementation Only certain augmented Inservice Inspection programs are included in the Fourth Ten Year Inspection Interval. These programs and their requirements are stated below. For information regarding other programs, contact the Plant Manager at the Kewaunee Nuclear Power Plant.

a. Nuclear Regulatory Commission Interim Head Inspection Requirements - NRC Order EA-03-009 Rev.1.

NRC Order EA-03-009 Rev. 1 summarizes requirements for performing Reactor Vessel Closure Head Inspections.

b. Nuclear Regulatory Commission NRC Bulletin 2003-02: Leakage From Reactor Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity.

NRC Bulletin- 2003-02 summarizes domestic PWR plant experiences where boric acid leakage has had the potential to degrade Reactor Pressure Vessel Lower Heads for in-core nuclear instrumentation. During the Fourth Inspection Interval, the 36 Bottom Mounted Instrumentation (BMI's) at the Kewaunee Nuclear Power Plant will receive a 100% Bare Metal VT-3 visual examination each Refueling Outage. The results of these inspections shall be documented in the ISI reports and reported in the Inservice Inspection Summary Report following the outage in which the examinations were conducted.

c. Nuclear Regulatory Commission NRC Bulletin 87-01 Thinning of Pipe Welds in Nuclear Power Plants and Nuclear Regulatory Commission Generic Letter 89-08 Erosion/Corrosion - Induced Pipe Wall Thinning.

Examinations of Flow Assisted Corrosion (FAC) is governed by Kewaunee Nuclear Power Plant FP-PE-FAC-01 Flow Accelerated Corrosion Inspection Program.

F-2 Rev. 1

Appendix F Augmented Examination Programs

d. Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Coolant Boundary Components in PWR Plants".

Generic Letter 88-05 summarizes domestic PWR plant experiences where boric acid leakage has had the potential to degrade carbon steel Reactor Coolant System pressure boundary components. The NRC staff noted that boric acid leakage potentially affecting the integrity of the reactor coolant pressure boundary should be procedurally controlled to ensure continued compliance with the licensing basis. KNPP currently administers one all encompassing formal program for the control and correction of boric acid leakage. However, in letter, K-88-205 from the NRC to KNPP in response to KNPP Letter NRC-88-77, the NRC noted that KNPP should maintain in auditable form, records of the Generic Letter 88-05 program and results obtained from implementation of the Generic Letter 88-05 program. To this end, in addition to the current Boric Acid Corrosion Control Inspection and Evaluation Program, KNPP identified, for the Generic Letter 88-05 Program, in the ISI Class 1 and Class 2 Boundaries, 39 valves, 4 flanges, Manways, Reactor Vessel Closure Head, Reactor Vessel Bottom Head and 72 Reactor Coolant Pump Bolts with carbon and low alloy steel where leaks that are smaller than the allowable Technical Specification limit could cause degradation of the primary pressure boundary by boric acid corrosion:

SI-13A SI-22B FE458 SI-4A RHR-2B PR-3B LD-2 SI-13B SI-303A FE-459 SI-4B RHR-11 PS-1A LD-3 SI-21A SI-303B SI-2A RHR-1A RC7103A PS-1B LD4A SI-21B SI-304A SI-2B RHR-1B RC-103B CVC-1l LD-4B SI-22A SI-304B SI-3 RHR-2A PR-3A CVC-15 LD-4C PR-1A PR-1B PR-2A PR-2B PR-F1 PR-F2 RC-402 RC-412 Reactor Vessel Closure Head - 33 Control Rod Drive Mechanisms, 3 Core Exit Thermocouple Nozzle Assemblies, 1- 1" Reactor Vessel Head Vent Line and 1 - 1" Reactor Vessel Level Instrumentation Line.

Reactor Vessel Bottom Head - 36 Bottom Mounted Instrumentation (BMI's).

Steam Generator Manway Bolting: SG-1A-HLMWB, SG-1A-CLMWB, SG-1B-

,HLMWB and SG-lB-CLMWB: 16 Studs, 16 Nuts and 32 Washers each Manway.

Pressurizer Manway Bolting P-MWB (P-B 1 thru P-B 16).

Reactor Coolant Pump Main Flange Bolting: RCP-B 1 through RCP-B48.

Reactor Coolant Pump No.1 Seal Housing Bolting: RCP-B49 through RCP-B72.

Note: Reactor Vessel Closure Head Studs, Nuts and Washers (48 each) are not included in F-3 Rev. 1

Appendix F Augmented Examination Programs the Generic Letter 88-05 Program due to cleaning and/or neolubing performed each 18 month Refueling Outage..

During the Fourth Inspection Interval, pressure retaining components of these valves, flanges, manways, Reactor Vessel Closure Head, Reactor Vessel Bottom Head and Reactor Coolant Pumps will receive a VT-3 visual examination each period (3 1/3 Years).

The results of these examinations shall be documented in the ISI reports and reported in the Inservice Inspection Summary Report following the outage in which the examinations were conducted.

e. Updated Safety Analysis Report, Section 4.2.2, Reactor Coolant Pump Flywheels Section 4.2.2 reports that the reactor coolant pump flywheels are designed in part to preclude missile generation by the pump flywheels. The design included a fracture mechanics evaluation of the reactor coolant pump flywheel. The evaluation considered the following assumptions:

i. Maximum tangential stress at an assumed over speed of 125%.

ii. A crack through the thickness of the flywheel at the bore.

iii. Four hundred (400) cycles of startup operation in forty years.

Using critical stress intensity factors and crack growth data attained on flywheel material, the critical crack size for failure was shown to be greater than 17 inches radially and the crack growth rate was 0.030 inch to 0.060 inch per 1000 cycles.

Periodic ultrasonic examinations will provide continued assurance that the, flywheels are structurally sound. The ultrasonic examination procedure utilized for these periodic examinations must be capable of detecting at least '/2-inch deep cracks from the ends of the flywheel. Flywheels on both reactor coolant pumps will be ultrasonically examined during the Fourth Inspection Interval. Examinations will be performed concurrent with scheduled maintenance.

f. Plant Technical Specification, TS 4.2.b, Steam Generator Tubes Examination of steam generator tubing is governed by TS 4.2.b. Refer to TS 4.2.b for details regarding this program.

g. Plant Technical Specification, TS 4.14, Testing and Surveillance of Shock Suppressors (Snubbers)

All safety-related hydraulic shock suppressors are visually examined and tested in accordance with the requirements of TS 4.14. Refer to TS 4.14 for details regarding this program.

F-4 Rev. 1

Appendix F Augmented Examination Programs

h. Kewaunee Nuclear Power Plant Alloy 600/82/182 Weld Program.

The following components/welds at the Kewaunee Nuclear Power Plant contain Alloy 600/82/182 material in the Reactor Coolant Pressure Boundary:

IDENTITY LOCATION MATERIAL

1. RV-P1 through RV-P36 Reactor Vessel Bottom Head "Bottom Alloy 600 Mounted Instrumentation's (BMI's)"

2. SI-W54DM Reactor Vessel 4" Nozzle Safe End Alloy 600 with Nozzle End of SA182

3. SI-Wi 12DM Reactor Vessel 4" Nozzle Safe End Alloy 600 with Nozzle End of SA182

4. RC-W76DM Steam Generator A 29" ID Nozzle Alloy 600 with Safe End Alloy 690 Cladding

5. RC-W77DM Steam Generator A 31" ID Nozzle Alloy 600 with Safe End Alloy 690 Cladding

6. RC-W78DM Steam Generator B 29" ID Nozzle Alloy 600 with Safe End Alloy 690 Cladding

7. RC-W79DM Steam Generator B 31" ID Nozzle Alloy 600 with Safe End Alloy 690 Cladding

8. PR-WIDM Pressurizer Relief 6" Nozzle Safe End SA216 Grade WCC Nozzle Casting with Clad Austenitic Stainless Steel fitted with a TP3 16L Stainless Steel Nozzle Safe End. Note: No Alloy 600/82/182 Product Form

9. PR-W16DM Pressurizer Safety 6" Nozzle Safe End SA216 Grade WCC Nozzle Casting with Clad Austenitic Stainless Steel fitted with a TP316L Stainless Steel Nozzle Safe End. Note: No Alloy 600/82/182 Product Form

10. PR-W26DM Pressurizer Safety 6" Nozzle Safe End SA216 Grade WCC Nozzle Casting with Clad Austenitic Stainless Steel fitted with a TP316L Stainless Steel Nozzle Safe End. Note: No Alloy 600/82/182 Product Form

11. PS-W61DM Pressurizer Spray 4" Nozzle Safe End SA216 Grade WCC Nozzle Casting with Clad Austenitic Stainless Steel fitted with a TP316L Stainless Steel Nozzle Safe End. Note: No Alloy 600/82/182 Product Form F-5 Rev. 1

Appendix F Augmented Examination Programs IDENTITY LOCATION MATERIAL

12. RC-W67DM Pressurizer Surge 14" Nozzle Safe End SA216 Grade WCC Nozzle Casting with Clad Austenitic Stainless Steel fitted with a TP316L Stainless Steel Nozzle Safe End. Note: No Alloy 600/82/182 Product Form During the Fourth Inspection Interval, the following pressure retaining components will receive a 100% Bare Metal VT-3 Examination each Refueling Outage.

1. Reactor Vessel Bottom Head "Bottom Mounted Instrumentation's (BMI's)" - RV-P1 through RV-P36 The results of these examinations shall be documented in the ISI reports and reported in the Inservice Inspection Summary Report following the outage in which the examinations were conducted.

During the Fourth Ten Year Inspection Interval, the following pressure retaining welds will receive a Liquid Penetrant and Ultrasonic Examination following Insulation Removal and a VT-2 Examination with the Insulation in place as required by and at the frequency stated in ASME Boiler and Pressure Vessel Code Section XI 1998 Edition 2000 Addenda.

1. SI-W54DM

2. SI-W I12DM

3. RC-W76DM

4. RC-W77DM

5. RC-W78DM

6. RC-W79DM The results of these examinations shall be documented in the ISI reports and reported in the Inservice Inspection Summary Report following the outage in which the examinations were conducted.

i. In 1979, Kewaunee NuclearPowerPlant discovered shallow cracks in the feedwaternozzle-to-pipe welds for both steam generators. Inspection and repair details are documented in WPSC letter dated July 26, 1979, from E. R. Mathews (WPSC) to J. G. Keppler (NRC). Following this incident, Kewaunee Nuclear Power Plant examined these areas on a routine basis. In 1995, the Kewaunee Nuclear Power Plant again discovered shallow cracks in the feedwater nozzle-to-pipe welds for both steam generators. Evaluation and Acceptance was documented in Westinghouse Electric Corporation WCAP-14359 "Structural Integrity Evaluation for the Feedwater Nozzle Safe End Region of the Kewaunee NuclearPowerPlant". Following this incident, Kewaunee Nuclear Power Plant examined these areas on a routine basis. During the Steam Generator Replacement in 2001 Kewaunee Nuclear Power Plant replaced the welds in the feedwater nozzle to pipe with a different design. In the Fourth Ten YearInspection Interval, KewauneeNuclearPowerPlantwill radiograph or ultrasonically examine the feedwater nozzle-to-pipe welds each period when practicable (i.e.,

when refueling outage schedule permits, to coincide with maintenance activities, etc.).

F-6 Rev. 1