ULNRC-05344, Response to Request for Additional Information Regarding 10 CFR 50.55a Request I3R-01 (Request for Relief from ASME Section XI Inservice Inspection Requirements for Third 10-Year Inspection Interval)

Response to Request for Additional Information Regarding 10 CFR 50.55a Request I3R-01 (Request for Relief from ASME Section XI Inservice Inspection Requirements for Third 10-Year Inspection Interval)

ML063460055
Person / Time
Site:	Callaway
Issue date:	11/29/2006
From:	Fitzgerald D AmerenUE
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-05344
Download: ML063460055 (19)

Text

AmerenUE PO Box 620 Callaway Plant Fulton, MO 65251 I

November 29, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 Ladies and Gentlemen: ULNRC-05344 10 CFR 50.55a V1Ameren UE DOCKET NUMBER 50-483 UNION ELECTRIC COMPANY CALLAWAY PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING 10 CFR 50.55a REQUEST 13R-01 (REQUEST FOR RELIEF FROM ASME SECTION XI INSERVICE INSPECTION REQUIREMENTS FOR THIRD 10-YEAR INSPECTION INTERVAL)

By letter dated March 28, 2006, Union Electric Company (AmerenUE) requested NRC approval of several relief requests (attached to the letter) for the third 10-year inservice inspection interval at Callaway, pursuant to 10 CFR 50.55a(a)(3) and/or 10CFR 50.55a(g)(5)(iii). As noted in the March 28 letter, the Code Edition (and Addenda) applicable to Callaway for its third inspection interval, which began December 19, 2005, is the ASME Boiler Pressure Vessel Code,Section XI, 1998 Edition through 2000 Addenda.

One of the submitted relief requests, 13R-0 1, which is still under review by the NRC staff, is a request to continue the application of an alternative risk-informed methodology for the inservice inspection of Class 1 and 2 piping welds at Callaway, as previously permitted under a 10 CFR 50.55a request that was approved by the NRC during Callaway's second ten-year inspection interval. AmerenUE requested approval of Relief Request 13R-01 by the end of this year, i.e., well in advance of the next refueling outage (scheduled for Spring 2007) since inspection activities planned for the outage are affected by the relief request.

Subsequent to the March 28, 2006 submittal, several questions / requests for additional information were identified by the NRC staff from their review of Relief Request 13R-03. Responses to these questions/requests have been prepared and are hereby provided as an attachment to this letter. Some of the questions/responses are a subsidiaryof Ameren Corporation

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ULNRC-05344 November 29, 2006 Page 2 related to the plant Probabilistic Risk Assessment (PRA) performed for Callaway, as they refer to a list of Facts and Observations (F&Os) that were identified from the peer review of Callaway's PRA. This list of Level 'A' and 'B' F&Os was electronically provided to the NRC staff for their information, but it is also included herein as an attachment to this letter. Finally, in responding to the NRC staff's questions/requests, the need to make some minor changes to the version of Relief Request 13R-01 (Rev. 0) that was submitted last March was identified. An updated version of Relief Request 13R-01 (Rev. 0) is therefore provided as an attachment to this letter.

It may be noted that no new regulatory commitments have been made or identified pursuant to this letter and its attachments. Please contact me at 573 676-8659 or Dave Shafer at 314-554-3104 for any questions you may have regarding the attached.

Sincerely, David T. Fitzgerald Manager - Regulatory Affairs TBE/jdg Attachments: Responses to NRC Questions / Requests for Information Level 'A' and 'B' F&Os for Callaway PRA Relief Request 13R-01, Revision 0

ULNRC-05344 November 29, 2006 Page 3 cc: U.S. Nuclear Regulatory Commission (Original and 1 copy)

Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 Mr. Bruce S. Mallett Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N. Donohew (2 copies)

Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 7E1 Washington, DC 20555-2738 Missouri Public Service Commission Governor Office Building 200 Madison Street PO Box 360 Jefferson City, MO 65102-0360

ULNRC-05344 November 29, 2006 Page 4 bcc: D. T. Fitzgerald (w/o)

G. A. Hughes (w/a)

D. E. Shafer (wla) (470) (2 copies)

S. L. Gallagher (w/o) (100)

C. J. Struttmann (NSRB)

K. A. Mills (w/a)

D. J. Maxwell (w/a)

G. A. Forster (w/a)

J. A. Doughty (w/a)

S. L. McCracken (w/a)

T. B. Elwood (w/a)

A160.0761 Chrono file The following are each provided a copy without attachments:

Ms. Diane Hooper Mr. Carl Corbin Supervisor, Licensing STARS Regulatory Affairs Manager WCNOC Comanche Peak SES P.O. Box 411 P.O. Box 1002 Burlington, KS 66839 Glen Rose, TX 76043 Mr. Scott Bauer Mr. Dennis Buschbaum Regulatory Affairs Comanche Peak SES Palo Verde NGS P.O. Box 1002 P.O. Box 52034, Glen Rose, TX 76043 Mail Station 7636 Phoenix, AZ 85072-2034 Mr. Scott Head Mr. Stan Ketelsen Supervisor, Licensing Manager, Regulatory Services South Texas Project NOC Pacific Gas & Electric Mail Code N5014 Mail Stop 104/5/536 P.O. Box 289 P.O. Box 56 Wadsworth, TX 77483 Avila Beach, CA 93424 Mr. John O'Neill Certrec Corporation Shaw, Pittman 4200 South Hulen, Suite 630 2300 N. Street N.W. Fort Worth, TX 76109 Washington, DC 20037

Responses to NRC Questions / Requests for Information Regarding 10 CFR 50.55a Request 13R-01 Initial Question/Request:

Discuss in detail the technical basis for including all alloy 600 pressure-retainingdissimilarmetal (DM) butt welds at operatingtemperaturesgreaterthan 450°Fin your risk-informed inservice inspection (RI-ISI) program, ratherthan a separateaugmented program in addition to the RI-ISI program.

Response

In response to the underlying concern behind this question, Relief Request 13R-01 has been revised to ensure that the subject examinations (for the alloy 600 pressure-retaining DM butt welds) will be performed in accordance with ASME Section XI requirements, including the need to submit a request(s) for relief for incomplete examinations.

Question 1:

In system BB (reactorcoolant), of those in risk category 2 with assigned damage mechanism TT (formerly in the ASME Class 1 Code Category B-J), Attachment 1 (page 5 of 7) indicates that during the "1st Approved RI-ISI Interval," of the 33 welds in this grouping, 8 were selected for examination. Yet, for this same grouping in the originalsubmittal of 2/16/2001, Table 5-2 on p 21 of 22 indicates that of the 33 welds in this grouping, 9 were selected for examination. Explain the apparent discrepancy.

Response

The actual number selected in the database for BB, Risk Category 2, with a DM of TT is 9. The number in the original table attached to 13R-01 (Rev. 0) was incorrect and has been corrected, as reflected in the updated version of 13R-01 (Rev. 0) provided.

Question 2:

In system EP (accumulatorsafety injection), of those in risk category 5 with no assigneddamage mechanism (formerly in the ASME Class 1 Code Category B-J), Attachment 1 (page 6 of 7) indicates that during the "1st Approved RI-ISI Interval," of the 12 welds in this grouping, 2 were selected for examination. Yet, for this same grouping in the originalsubmittal of 2/16/2001, Table 5-2 on p 22 of 22 indicates that of the 12 welds in this grouping, only 1 was selected for examination. Again, explain the apparentdiscrepancy.

Response

One weld was added to the examination scope shortly following implementation of RI-ISI. This may have been done to raise the percentage for the EP system to more closely match what was identified for Wolf Creek (WCNOC).

Question 3:

In system BB (reactorcoolant), of those in risk category 6 with assigneddamage mechanism TT (formerly in the ASME Class 1 Code Category B-J, Attachment 1 (page 5 of 7) indicates a populationof 30 welds during the "1st Approved RI-ISI Interval." However, in the original submittal of 2/16/2001, Table 5-1 on p 21 of 22 indicates for these same 30 welds that 18 of them are in Category 6, and 12 of them are in Category 7, the difference being that the latter 12 were found to have low consequence ratherthan medium consequence ranking. Again, explain the apparentdiscrepancy.

Response

Immediately following the submittal in February 2001, during discussions with WCNOC regarding the consequence of the RCP seal injection line failing, it was decided - as a conservative measure - to change the applicable segments from risk category 7 to risk category 6 for consistency with WCNOC. This was not considered to be a significant change (because it did not affect examination requirements or percentages) and therefore a follow-up submittal was not made to note this change.

Question 4:

In the original submittal of 2/16/2001, the licensee indicated that "additional examinations will be performed on these elements up to a number equivalent to the number of elements required to be inspected on the segment or segments initially. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined." RAIs on this aspect of the RI-ISI program addressed the scope (i.e., the number) of additional examinations and the method of selection of locations for these examinations.

However, the timeframe for completing these additional examinations, particularly if flaws are found in the first sample expansion, was not specified. Please provide a timeframe for completing all required additional examinations. The staff expects that sample expansion examinations will be performed in the same time frame that is outlined in the ASME,Section XI IWB-2430.

Response

If the initial examinations scheduled for an outage reveal the presence of flaws such that additional examinations are required, the additional examinations will be performed during that same (i.e., the current) outage, as specified in IWB-2430.

Question 5:

In the originalsubmittal of 2/16/2001, the licensee indicated that "A Westinghouse Owners Group (WOG) PRA Peer Review of the Callaway PRA was performed in November of 2000. This review was performed on model PRAUPDT2, which is the second Callaway PRA update. This second update featured changes to loss of coolant accident (LOCA) initiating frequencies, which were updated to current industry values, and several other modeling enhancements. Both the review and the update to the PRA model occurred after the RI-ISI project was completed.

However, the LOCA frequencies and importantmodeling changes were taken into account during the RI-ISI project," and [it was noted] that "thepreliminary finding of the WOG PRA Peer Review indicates that all PRA elements were gradedas sufficient to supportmeaningful rankings for the assessment of systems, structures,and components, when combined with deterministicinsights."

While the staff acknowledges that the licensee took LOCA frequency changes and important modeling changes into account during the originalRI-ISI project, Regulatory Guide (RG) 1.178,

'An Approach for Plant-Specific Risk-Informed Decisionmakingfor Inservice Inspection of Piping,"

Revision 1, dated September2003, includes guidance on what should be included in risk-informed inservice inspection (RI-ISI) submittals, particularlyin dealing with probabilisticrisk assessment (PRA) issues. Specifically, on page 28 of RG 1.178, the following is stated regarding the information that should be included in a submittal:

"A description of the staff and industry reviews performed on the PRA limitations, weakness, or improvements identified by the reviewers that could change the results of the PRA should be discussed. The resolution of the reviewer comments, or an explanation of the insensitivity of the analysis used to support the submittal to the comment, should be provided."

During a conference call with the licensee on August 8, 2006, the licensee indicated that all of the Level A and B Facts and Observations (F&Os) from the above PeerReview have been addressedand incorporatedinto the PRA model that was used to re-perform the RI-ISI analyses in preparationfor renewing the program.Please confirm this and provide a listing of these Level A and B F&Os, along with their resolutions.

In addition, please identify any other "openitems" with the PRA model that was used to re-perform the RI-ISI analyses in preparationfor renewing the program that would meet the thresholdof a Level A or B F&O, and explain why resolving the open item would not have a potentially significant impact on the RI-ISI program (eitherfrom the risk-significanceof pipe segments or from an overall delta-riskperspective).

Response

The following bulleted points clarify Callaway's response to the above questions/requests as discussed during the conference call on August 8, 2006:

" The RI-ISI analyses performed in preparation for renewing the program were performed with the third PRA Update model revision. This model revision incorporated some of the responses to the significance Level A and B WOG Peer Review Facts and Observations (F&Os), but not all.

• The Fourth PRA Update model revision was completed in April, 2006 and was primarily undertaken to generate a model that would satisfy the quality requirements for the MSPI program. As such, all but five of the A and B F&Os were addressed and incorporated into the Fourth PRA Update model revision. The five that were not addressed did not impact the MSPI program and are identified in Callaway's MSPI Basis Document.

" An evaluation of the RI-ISI analyses was performed using the Fourth PRA Update model revision. This was done as part of procedural guidance that requires review of past PRA applications after a PRA model update. The evaluation determined that no consequence segments had their ranks increase due to the Fourth PRA Update model revision. Section 4.2 of NEI 04-05, "Living Program Guidance to Maintain Risk-Informed Inservice Inspection Programs for Nuclear Plant Piping Systems," states that "for the EPRI RI-ISI methodology, as long as the consequence rank assignments are consistent between the original PRA and the updated PRA.... then these results can be documented and no further analysis is required." Thus, the Fourth PRA Update model revision did not change the RI-ISI results.

• The five A and B F&Os that remain open are identified in the attached listing. It has been determined that these five F&Os would have no appreciable impact on the RI-ISI analyses.

The requested listing of A and B F&Os and their resolutions is provided as an attachment.

It may be noted that the only Significance Level C F&O of potential significance is item AS-5.

This F&O pertains to tracking the resolution of the WOG RCP seal LOCA model issue. The Callaway PRA uses a seal LOCA model based on WCAP-1 0541. The final resolution of the WOG RCP seal LOCA issue was issuance of the WOG2000 RCP seal LOCA model.

Callaway plans to implement the WOG2000 model into the next PRA update. A cursory review of the differences between the two models indicates that the WOG2000 RCP seal LOCA model will not appreciably affect the RI-ISI analyses.

Question 6:

In the list of Facts and Observations(F&O's) documented from the peer review of the licensee's plant PRA (as provided in response to Question 5 above), F&O item L2-1 is identified as "open" with respect to its current status. F&O item L2-1 addresses failure of containment isolationand internalfloods in the licensee's LERF calculation. Although the licensee has stated that this F&O item would not appreciablyaffect the RI-ISI LERF evaluation, more detailedor furtherjustification should be provided for this conclusion.

Response

Yes, WOG PRA Peer Review Fact & Observation (F&O) L2-1 (Significance Level A) remains open. This F&O pertains to the need to address failure of containment isolation and internal flooding in the Large Early Release Frequency (LERF) calculation.

The containment isolation issue had actually been addressed during the generation of the LERF calculation by the fact that failure of containment isolation is a very low probability event. The Peer Review reviewers agreed but stated that it should be included for applications in which failure of containment isolation was likely to occur.

The calculation of core damage frequency (CDF) due to internal flooding was a stand-alone calculation (apart from the Level 1 internal events CDF calculation). Therefore flooding was originally not included in the calculation of LERF, which was based on the Level 1 internal events CDF calculation.

The impacts due to failure of containment isolation and flooding were evaluated for both CDF and LERF in the Risk Informed ISI consequence evaluation. If a containment isolation valve was impacted by the assumed weld rupture (due to spray effects, flooding effects, impingement effects, environmental effects, etc.), the Conditional Large Early Release Probability (CLERP) was set equal to the Conditional Core Damage Probability (CCDP). Flooding effects (spray, impingement, submersion, etc.) due to the assumed weld rupture were evaluated for impact on all equipment in a given flood area. Equipment potentially impacted was assumed to fail for the calculation of CCDP and CLERP for that weld.

Because the risk-informed ISI consequence evaluation included the impacts due to failure of containment isolation and flooding on both CDF and LERF, the final resolution of F&O L2-1 will result in no changes in consequence rankings.

F and 0 List and Status F Significance Status F&O Desription~ ~2~ CommentsJ Utilize the distribution in an industry publication (such as The LOSP and SGTR frequencies were updated and the small NUREGICR-5750) and Bayesian update the LOSP initiating event LOCA frequency was evaluated.

IE-4 B Closed frequency generic distribution. Develop a small LOCA frequency based on RCP seal LOCA initiating events and piping failures. Use industry data to develop a distribution for SGTR.

ISLOCA is not a significant contributor to core damage frequency (CDF), comprising only about one percent of the Callaway CDF. In Use recent guidance (such as NUREG/CR-5102) on interfacing regards to large early release frequency (LERF), LERF was system LOCAs to gauge the effect or revise the analysis. conservatively evaluated per the EPRI methodology for the RI-ISI application. This F&O would not appreciably affect the RI-ISI LERF evaluation.

Add consideration for CCW restoration prior to the H3 top event in Event tree modified.

the Loss of All CCW event tree.

Include consideration in the sequence 2 branch of the Loss of All Event tree modified.

AS-2 B Closed Service Water event tree for core uncovery probability at the time of service water recovery.

AS-4 B Closed Require SG integrity for sequences that credit secondary side heat Event trees modified, as well as steam and feedwater isolation fault removal in events TMSI and TMSO trees.

The guidance used for the recovery analyses for CCW and SW The Human Reliability Analysis (HRA) was updated, using current should be obtained, documented, and used. That guidance does not methods, for risk significant human interactions in order to satisfy have wide application in the industry, so independent review may be the requirements of ASME Capability Category II. The human warranted. With respect to the complexity judgment in the interactions associated with this F&O are risk-insignificant.

calculation, actual experience should be reviewed and the judgments reviewed by plant personnel.

Review the basis for determining that the 125VDC equipment can Basis was Human reviewed Reliability and a more Analysis (HRA)thorough justification was updated, using provided.

current The and HuaReibltAnyss(A)wspdeuigcret reliably surve loss proeliaby survive of room losroomh oling for cooling discoss n on-S cases, non-LOSP forif castes PRA methods, for risk significant human interactions in order to satisfy provide a more thorough discussion/justification in the PRA the requirements of ASME Capability Category II. The human documentation. Also provide a more thorough discussion of the interaction associated with this F&O is risk significant based on operator action, timing, and cues for the action to open the room Fussell-Vesely (F-V), and was therefore redone during the HRA TH-1 B Closed doors. In addition, consider performing sensitivity studies to update.

determine the effect on the PRA results of failure of the 125VDC switchgear #1 and #4 due to loss of cooling; the results of this should either support the basis for not requiring modeling of loss of cooling or indicate that there is a sensitivity that should be addressed in the discussion of uncertainties in the PRA results, and possibly be examined explicitly for certain risk-informed applications.

Page 1 of 4

F and 0 List and Status Clarify the timing bases for each human action modeled. Clearly The Human Reliability Analysis (HRA) was updated, using current TH-2 B Closed reference an applicable analytical basis for each, and explain methods, for risk significant human interactions in order to satisfy in thethreuemnsoASECpblyCagryI.Teiigbss TH-2 B Clsed HRA notebook what attributes, assumptions, and specific results of the requirements of ASME Capability Category 11.The timing basis thR refenotebk watariues, assumptthtions, ansfor these human interactions was clarified and referenced.

the referenced analyses support the timing.

Consider preparing success criteria guidance for the PRA, to The issue described by this F&O is primarily a documentation issue address such items as overall success criteria definition process, and therefore has no impact on RI-ISI. It should also be noted that development of success criteria for systems, human actions, and the Callaway PRA success criteria are very similar to those of other, TH-3 Open sequences, application of bounding versus best-estimate like-design plants.

assumptions in analyses, and limitations of analysis codes. Also consider creating a clear, traceable path from the event sequences to the supporting success criteria and to the underlying analytical basis for the criteria.

Review the MAAP analyses made using the earlier version of MAAP The MAAP analyses were reviewed, and it was determined that to see if any such analyses were used to support current success results would not have been significantly affected (i.e., success TH-4 B Closed criteria, and, if so, determine whether or not the results would be criteria would not have changed nor would human interaction timing affected if the corrected version of the code (with the enhanced have changed such that human error probabilities would have been pressurizer model) were used. Or, make use of other transient adversely affected).

analysis results for such cases.

Since the MAAP 3.0B code is generally not accepted as providing Other transient analysis results were used in an evaluation to show valid results for the early blowdown phases of large breaks, consider that the accumulator success criteria are justified. A sensitivity dselecting another means of demonstrating the required number of analysis showed that the Callaway CDF is insensitive to the TH-7 CoseB seectng noter ean of emostrtin th reuird nmbe of accumulator success criteria.

accumulators; testing the sensitivity of the PRA results to the accumulator success criterion; and/or providing additional justification for the selection of the existing success criterion.

Add separate common cause failure events for the "failure to run" Separate common cause failure basic events were added to all SY-1 A Closed failure mode for applicable component groups (i.e., pumps, diesel affected fault trees.

_generators, fans, etc.).

Clarify the assumptions and modeling of ESW operation to supply Assumptions and modeling clarified and fault trees modified.

SY-7 B Closed AFW and revise the low suction pressure (LSA and LSB) fault tree logic.

Revise system fault tree logic (loss of RCP seal cooling, train B Fault tree modified.

SY-9 A Closed CCW) to ensure train B CCW heat exchanger TM event is properly accounted for during the quantification process.

DA-1 B Closed Perform an update using recent plant component failure rates. Risk-informed data update performed.

Provide guidance and the use of the CRPS, perhaps performing a Data update guidance provided, and utilized during data update.

DA-2 B Closed comparison with a lognormal update.

Provide a grouping process of component types that accounts for Data update grouped components by common cause groupings.

DA-3 B Closed such items as size, service conditions, frequency of demands, and environmental conditions.

Page 2 of 4

F and 0 List and Status ethe common cause failure probabilities using more recent Common cause failure probabilities were updated using a more Updaterecent source.

DA-4 B Closed sources. The INEEL data base is an on-going program that represents the largest compendium of common cause events.

The Human Reliability Analysis (HRA) was updated, using current Use of screening values for pre-initiator errors and simple actions methods, for risk significant human interactions in order to satisfy HR-1 B Closed was adequate for the IPE, but a PRA intended for risk-informed the requirements of ASME Capability Category I1. The risk applications should use plant-specific values for HEPs. significant human interactions associated with this F&O are now plant-specific.

The Human Reliability Analysis (HRA) was updated, using current methods, for risk significant human interactions in order to satisfy HR-2 B Closed Process guidance for the HRA was found, however it did not seem the requirements of ASME Capability Category I1. A consistent HRA to have been followed in the development of the HRA. approach, including the selection of current methods, was used in the HRA update.

The Human Reliability Analysis (HRA) was updated, using current methods, for risk significant human interactions in order to satisfy HR-4 B Closed There did not appear to be a systematic process for labeling of the requirements of ASME Capability Category I1. The identification human interactions and relating them to procedural guidance. and classification of all post-initiator human failure events was re-visited and updated as part of this update.

The development of execution error rates is not substantiated, The Human Reliability Analysis (HRA) was updated, using current Tdocumepmentor exeproducutionlerror ralthouhthes notis san , methods, for risk significant human interactions in order to satisfy HR-5 B Closed documented, or reproducible, although the method is clearly the requirements of ASME Capability Category II. The execution discussed and is consistent with industry standards for 1992 (time error rates for the risk significant human interactions were updated when the analysis was performed). as part of this update.

The Human Reliability Analysis (HRA) was updated, using current The assignment of the cognitive error is not based on a generic methods, for risk significant human interactions in order to satisfy HR-7 B Closed industry practice. It was based on an un-referenced method the requirements of ASME Capability Category II. The cognitive developed by PP&L. error rates for the risk significant human interactions were updated as part of this update.

The Human Reliability Analysis (HRA) was updated, using current methods, for risk significant human interactions in order to satisfy HR-8 B Closed The quantification process did not consider the sequence context of the requirements of ASME Capability Category I1. This update the HEPs. considered plant-specific inputs, procedures, and performance shaping factors as well as PRA scenario-specific factors in the definition and quantification of the HEPs.

In the original IPE, a post-quantification search for dependent human The Human Reliability Analysis (HRA) was updated, using current methods, for risk significant human interactions in order to satisfy HR-9 A Closed errorsnotindocumented was the same cutset was nor is performed.

it obvious Unfortunately, this effort this resulted in theeffort promdadpnec the requirements of ASME nlssadfudsvrldpnece Capability Category II. This update assignment of any dependent action HEPs. performed a dependency analysis and found several dependencies that were incorporated.

Page 3 of 4

F and 0 List and Status F Significance Status' F& Description Comments The Human Reliability Analysis (HRA) was updated, using current methods, for risk significant human interactions in order to satisfy HR-10 B Closed There were several cutsets with multiple HEPs that seemed to be the requirements of ASME Capability Category I1. This update performed a dependency analysis that addressed this F&O issue.

ISLOCA is not a significant contributor to core damage frequency Consider updating the ISLOCA analysis to reflect approaches (CDF), comprising only about one percent of the Callaway CDF. In ST-1 B Open defined 5744 andin related references such as NUREG/CR-5102 and NUREG/CR-reports. regards to large early release frequency (LERF), LERF was conservatively evaluated per the EPRI methodology for the RI-ISI application. This F&O would not appreciably affect the RI-ISI LERF evaluation.

Formalize a process whereby at least some sample of non-dominant Process was formalized and utilized in the most recent PRA update.

cutsets are reviewed.

Containment isolation failure is not a significant contributor to large early release frequency (LERF), comprising less than 0.1% of LERF.

Address failure of containment isolation and Internal Floods in the LERF was conservatively evaluated per the EPRI methodology for L2-1 A Open LERF calculation, the RI-ISI application. In addition, spatial affects were evaluated such that if containment isolation were impacted, containment bypass was assumed. This F&O would not appreciably affect the RI ISI LERF evaluation.

L2-2 B Closed Reevaluate CLERP values. Guidance provided to reevaluate CLERP values during LERF updates.

LERF was conservatively evaluated per the EPRI methodology for the RI-ISI application. Most split fractions used in the containment Develop a process for applications to ensure the effects of plant event tree are based on expert solicitation and/or phenomenological L2-3 B Open changes on LERF containment event tree split fractions is considerations, and thus are not impacted by plant changes. There adequately characterized. are no known plant changes that would impact the split fractions utilized. This F&O would not appreciably affect the RI-ISI LERF evaluation.

Consider adding clarification to the PRA guidance to indicate that Procedural guidance provided to review PRA applications as part of MU-4 B Closed particular PRA applications may be excluded from the list of a PRA update and to document the review.

applications to be updated if appropriate rationale are provided as part of the PRA update (or other documentation).

Page 4 of 4

10 CFR 50.55a Request Number 13R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Components Affected All Code Class I and 2 piping welds previously subject to the requirements of ASME Section XI, Table IWB-2500-1 (Examination Categories B-F and B-J) and Table IWC-2500-1 (Examination Categories C-F-1 and C-F-2).

2. Applicable Code Edition and Addenda ASME Boiler and Pressure Vessel Code,Section XI 1998 Edition through 2000 Addenda

3. Applicable Code Requirement ASME Section XI, Tables IWB-2500-1 and IWC-2500-1 for Examination Categories B-F, B-J, C-F- 1 and C-F-2 stipulate the selection and examination requirements for Class 1 and 2 piping welds.

4. Reason for Request ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 originally contained the requirements for the nondestructive examination of Class I and 2 piping welds. In 2001, a risk-informed methodology for the inservice inspection of Class I and 2 piping welds was applied at the Callaway Nuclear Power Plant. The risk-informed inservice inspection process used in this application is described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Rev. B-A "Revised Risk-Informed Inservice Inspection Evaluation Procedure." The RI-ISI application was also conducted in a manner consistent with ASME Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B."

This risk-informed application met the intent and principles of Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" and Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping".

The original RI-ISI template, "Risk-Informed Inservice Inspection Program Plan - Callaway Plant (Revision 1)" was submitted to the NRC for approval per IOCFR50.55a(a)(3)(i) in Letter No.

ULNRC-4392, dated February 16, 2001. Based upon the information provided in the RI-ISI template, the request to implement the RI-ISI methodology on Class 1 and 2 piping welds was approved by the NRC in a letter dated January 30, 2002 (TAC No. MB11205). The purpose of this current request is for the continued application of the RI-ISI methodology on Class I and 2 piping welds during the third ISI interval based on the alternative providing an acceptable level of quality and safety.

RR 13R-0I Rev 0a3 (with Changes Accepted) I of 7

5. Proposed Alternative and Basis for Use The proposed alternative is to continue applying the Risk-Informed ISI criteria of EPRI TR-112657 during the third ISI interval in lieu of the requirements of ASME Section XI, Table IWB-2500-1 (Examination Categories B-F and B-J) and Table IWC-2500-1 (Examination Categories C-F-1 and C-F-2).

When Callaway submitted their initial RI-ISI application to the NRC for approval, the following standard clause was included in the template submittal:

"The RI-ISI program is a living program requiring feedback of new relevant information to ensure the appropriate identification of high safety significant piping locations. As a minimum, risk ranking of piping segments will be reviewed and adjusted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant specific feedback."

Most U.S. nuclear power plants have implemented RI-ISI Programs with this standard clause to perform periodic reviews and updates. To address this issue, NEI 04-05, "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems" has been developed. As part of the ISI Program Update for the third ISI interval at Callaway, a RI-ISI Living Program Evaluation was performed in accordance with NEI 04-05.

The objective of this evaluation was to review plant and industry activities that could impact the bases of the Callaway RI-ISI application as it enters the third ISI interval.

In accordance with NEI 04-05, the following aspects were considered during the evaluation:

• Plant Examination Results

" Piping Failures

- Plant Specific Failures

- Industry Failures

" PRA Updates

" Plant Design Changes

- Physical Changes

- Programmatic Changes

- Procedural Changes

• Changes in Postulated Conditions

- Physical Conditions

- Programmatic Conditions RR 13R-01 Rev 0a3 (with Changes Accepted) 2 of 7

The RI-ISI Living Program evaluation resulted in the following seven issues being addressed in the RI-ISI application:

• The plant changes made per Modification Package MP 01-1019 and further documented in CAR 200200055 were incorporated into the RI-ISI Program. This modification replaced a troublesome safety injection valve with a new one slightly upstream.

" A steam generator replacement modification had been completed since the initial RI-ISI application. This modification was primarily a "like-for-like" replacement and as such had minimal impact on the RI-ISI application. Minor updates were made to the RI-ISI documents to account for the replacement, deletion and addition of welds associated with the modification.

" Changes to the plant PRA resulted in changes to the consequence rankings and subsequent revisions to the RI-ISI Program for three consequence segments. These changes were all on Low Risk segments which remained Low Risk after the changes were incorporated. As such, there was a negligible impact on the overall RI-ISI application.

" The Class 2, 4" NPS auxiliary feedwater lines from the outboard isolation valve to the connection to the main feedwater piping in all four trains were added to the RI-ISI Program.

This resulted from a change in ASME Section XI Code criteria when updating to the 1998 Edition with 2000 Addenda wherein Class 2, 4" NPS and smaller auxiliary feedwater piping is no longer exempt.

" The Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP) values for the Class 2, 4" auxiliary feedwater piping addressed above were higher than the upper bound values previously used in the risk impact analysis. New upper bound values of 2.4E-2 and 2.4E-3 were used in the updated risk impact analysis for CCDP and CLERP, respectively.

" Class 2 Borated Refueling Water (BN) system piping greater than 4" NPS shown on Callaway Drawing No. ISI-M-22BN01 (Q) was considered to be exempt during Interval 2, but is being considered as non-exempt piping during Interval 3 to be consistent with the Wolf Creek Nuclear Generating Station. This will have minimal impact on the RI-ISI Program as this piping was conservatively evaluated as non-exempt piping during the initial RI-ISI application to remain consistent with the other (Strategic Teaming & Resource Sharing)

STARS utilities' RI-ISI submittals. As such, the RI-ISI bases have already been established for the BN system.

• Based on ongoing industry experience with Primary Water Stress Corrosion Cracking (PWSCC), Assumption No. 7 in the Degradation Mechanism Evaluation was deleted. This assumption had stated the following:

"Bi-metallic welds with Inconel buttering are not considered susceptible to the PWSCC degradation mechanism."

For the third interval, Callaway will comply with the selection criteria and frequency of MRP-139 for welds that are potentially susceptible to PWSCC. At Callaway, this consists of 14 welds where piping attaches to the reactor pressure vessel and pressurizer. The Callaway PWSCC Augmented Program is independent from the RI-ISI Program, yet these 14 welds are RR 13R-01 Rev 0a3 (with Changes Accepted) 3 of 7

in the scope of both programs. From both a technical and administrative standpoint, precedence for the examination of the 14 welds that are potentially susceptible to PWSCC will be taken from how RI-ISI Programs at BWRs interface with the NRC mandated program to examine welds that are potentially susceptible to Intergranular Stress Corrosion Cracking (IGSCC). NDE requirements, including qualification, examination coverage calculation, detection and sizing used for examination of these welds will be in accordance with the 1998 Edition with 2000 Addenda of ASME Section XI (i.e., PDI Quality Examinations) and incomplete examinations shall require a relief request in accordance with 10CFR50.55a.

During the update of the Callaway ISI Program in preparation for their third ISI interval, other minor corrections were identified (e.g., correction of weld numbers) and were evaluated as part of the RI-ISI Living Program Update. These had no impact on the RI-ISI Program beyond requiring minor editorial corrections to the RI-ISI Program documents.

The RI-ISI Program was reevaluated for the seven issues and other minor corrections using the applicable portions of the same risk-informed process that originally established the risk-informed inspection program. The reevaluation was performed by inserting the new information at the appropriate levels of the analysis. All of the cases that were evaluated in the risk impact analysis during the original RI-ISI application were reevaluated using the new information that was determined for the current application. Results of the risk impact reanalysis were that the overall plant risk as measured as a change in Core Damage Frequency and Large Early Release Frequency was further decreased as a result of the application of the new information. As such, the RMSI application on Class I and 2 piping welds still maintains an acceptable level of quality and safety.

A summary table of the welds in the RI-ISI Program along with any changes resulting from the issues addressed above is provided in Attachment 1.

6. Duration of Proposed Alternative This IOCFR50.55a Request is being proposed for use during the third inspection interval that begins on December 19, 2004 and ends on December 18, 2014.

7. Precedents The proposed alternative in this IOCFR50.55a Request was included in second interval Relief Request ISI-23 for the Callaway Plant. This Relief Request was submitted to the NRC for approval per IOCFR50.55a(a)(3)(i) in Letter No. ULNRC-4392, dated February 16, 2001. Based upon the information provided in the RI-ISI template, the request to implement the RI-ISI methodology on Class I and 2 piping welds was approved by the NRC in a letter dated January 30, 2002 (TAC No. MB 1205).

Resubmittal of a R.-ISI application on Class 1 and 2 piping welds has been conducted by the V.C. Summer Nuclear Station for their third ISI interval in SCE&G Letter No. RC-04-0148, dated September 8, 2004. The NRC review of this V.C. Summer IOCFR50.55a Request is currently ongoing.

RR 13R-0I Rev 0a3 (with Changes Accepted) 4 of 7

Attachment 1 Inspection Location Selection Comparison Between ASME Section XI Code and EPRI TR-112657 by Risk Category Risk Failure Potential V*'Approved RI-ISI Interval New RI-ISI Interval Systemt1 1r Consequence 1 Code Category Weld RI-ISI Other 2 Weld RI-ISI Other( 2 )

Category Rank Rank DMs RanI Count Count AB 6 Low Medium None Low C-F-2 152 0 157 0 AE 2 High High TASCS Medium C-F-2 V(3) 0(3) 9(3) 6(3)

AE 2 High High TT Medium C-F-2 00) 0(3) 80) 2(3)

AE 4 Medium High None Low C-F-2 0(3) 0(3) 6 (3) 1(3)

AE 5 (3) Medium (High) Medium TASCS, (FAC) Medium (High) C-F-2 17 2 18 2 C-F-I 16 0 16 0 AE 6 (3) Low (High) Medium None (FAC) Low (High) CF2 85 0 90 0 AL 6 Low Medium None Low C-F-2 0(3) 0(3) 126(3) 0(3)

BB 2 High High TASCS, TT Medium B-1 9 3 9 3 BB 2 High High TASCS Medium B-J 24 6 24 6 BB 2(2) High (High) High TT, (PWSCC) Medium (Medium) B-F 00) 0(4) 1(4) 0(4) 1(4)

B-F 10) 0(4) 0(4) 00)

BB 2 High High TT Medium ___393_

B-J 33 9 33 9 BB 4(2) Medium (High) High None, (PWSCC) Low (Medium) B-F 0(4) 03(4) 554) 13(4) 2 1(4) 5(4) 8(4) 0(4)

B-F BB 4 Medium High None Low B322343 B-i 325 29 342 34 BB 5 Medium Medium TT Medium B-A 2 I 2 1 BB 6 Low Medium None Low B-J 30 0 30 0 BG 2 High High TT Medium B-J 3 I 3 1 B-J 5 0 5 0 BG 4 Medium High None Low C- F-I 84 9 84 9 BG 5 Medium Medium TT Medium B-J 4 1 4 1 B-J I 0 I 0 BG 6 Low Medium None Low C-F-I 40 0 40 0 BG 7 Low Low None Low C-F-I1 14 0 140 RR 13R-0I Rev 0a3 (with Changes Accepted) 5 of 7

Attachment 1 Inspection Location Selection Comparison Between ASME Section XI Code and EPRI TR-112657 by Risk Category Systemt~t Risk Consequence 1 Failure Potential t Code Category I" WeldApproved RI-ISI IWl Interval New RI-ISI Interval 2)

WlRank o RI-ISI Othert Weld RI-ISI Other(2)

Category T__ Rank Ran Rank Rank____

DMs Count I _ _ Count BN 4 Medium High None Low C-F-i 4 0 4 0 BN 6 Low Medium None Low C-F-I 5 0 I11 0 EF 5 Medium Medium MIC, PIT Medium C-F-2 4 1 4 1 EF 6 Low Medium None Low C-F-2 46 0 46 0 B-J 19 2 19 2 4 Medium High None Low EJ C-F-I 431 44 431 44 EJ 7 Low Low None Low C-F- 1 35 0 35 0 EM 5 Medium Medium IGSCC Medium B-J 20 2 19 2 B-i 110 0 107 0 Low Medium None Low EM 6 C-F-I 206 0 206 0 EM 7 Low Low None Low C-F- I 24 0 24 0 EN 6 Low Medium None Low C-F-I 86 0 86 0 EP 5 Medium Medium IGSCC Medium B-J 12 I 12 2 EP 6 Low Medium None Low B-J 87 0 93 0 Notes:

1. System designations are as follows:

AB - Main Steam System AE - Main Feedwater System AL- Auxiliary Feedwater System BB - Reactor Coolant System BG - Chemical and Volume Control System BN - Borated Refueling Water Storage System EF - Essential Service Water System EJ - Residual Heat Removal System EM - High Pressure Coolant Injection System EN - Containment Spray System EP - Accumulator Safety Injection System RR 13R-0I Rev 0a3 (with Changes Accepted) 6 of 7

Notes (con't):

2. The column labeled "Other" is generally used to identify augmented inspection program locations that are credited beyond those locations selected per the RI-ISI process, as addressed in Section 3.6.5 of EPRI TR-l 12657.

3. Due to a change in ASME Section Xl Code criteria, 4" NPS Class 2 auxiliary feedwater piping was added to the ISI Program, and therefore the RI-ISI Program, for the first time during the third ISI interval. This consisted of Class 2 piping from the outboard isolation valve to the first check valve (i.e., system "AL") and piping from the first check valve to the branch connection to feedwater (i.e., system "AE") in all four trains. This piping and its associated weldnients were outside the scope of the original RI-ISI application.

4. Changes to the information shown for former Code Category B-F welds reflect the implementation of Callaway's PWSCC augmented program.

RR 13R-0I Rev 0a3 (with Changes Accepted) 7 of 7