RS-13-147, Additional Information Supporting Fifth Inservice Inspection Interval Relief Requests I5R-01, I5R-02, and I5R-06

Additional Information Supporting Fifth Inservice Inspection Interval Relief Requests I5R-01, I5R-02, and I5R-06

ML13151A107
Person / Time
Site:	Quad Cities
Issue date:	05/30/2013
From:	Simpson P Exelon Generation Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-13-147
Download: ML13151A107 (15)

Text

4300 Winfield Road Warrenville. IL 60555 Exelon Generation. 630 657 2000 Office RS 147 10 CFR 50.55a May 30, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Additional Information Supporting Fifth Inservice Inspection Interval Relief Requests 15R-01, 15R-02, and 15R-06

References:

1. Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, "Quad Cities Nuclear Power Station, Units 1 and 2, Fifth Interval Inservice Inspection Program Plan and Relief Requests," dated September 28, 2012

2. Email from B. Purnell (U.S. NRC) to K. Nicely (Exelon Generation Company, LLC), "Quad Cities Relief Request Nos. 15R -01, 15R-02, and 15R Request for Additional Information," dated April 24, 2013 (ADAMS Accession No. ML13115A232)

3. Email from B. Purnell (U.S. NRC) to K. Nicely (Exelon Generation Company, LLC), "Quad Cities Relief Request No. I5R Request for Additional Information," dated May 2, 2013 (ADAMS Accession No. ML13123A086)

In Reference 1, Exelon Generation Company, LLC (EGC) submitted relief requests associated with the fifth inservice inspection (ISI) interval Quad Cities Nuclear Power Station, Units 1 and 2.

The NRC requested additional information that is needed to complete the review in References 2 and 3. The attachments provide the requested information.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

Patrick R. Simpson Manager - Licensing

May 30, 2013 U.S. Nuclear Regulatory Commission Page 2 Attachments:

1. Response to NRC Request for Additional Information

2. Quad Cities Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection Summary Tables cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Quad Cities Nuclear Power Station

ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request I5R-01-1 Discuss in the relief request why the proposed alternative provides reasonable assurance of structural integrity or leak tightness of the subject components.

Response

The proposed alternative verifies structural integrity and leak tightness by ensuring no leakage at nominal operating pressure during the Class 1 System Leak test. This test is a Code required exam that verifies no through wall leakage for the entire Class 1 boundary. It is performed prior to start-up and unit operation every outage to provide confidence there is no leakage. The Standby Liquid Control nozzle is included as part of this test. Passing this test with no identified through wall leakage at the nozzle provides reasonable assurance of structural integrity and leak tightness of the Standby Liquid Control nozzle.

During normal plant operation, the Standby Liquid Control nozzle is inaccessible for direct visual observation based on its location in the drywell. This does not preclude the nozzle from being indirectly monitored. Leakage from the nozzle would be collected in the drywell sumps, prompting action from the station. No leakage from this nozzle has ever been observed at Quad Cities Nuclear Power Station (QCNPS) from this location.

NRC Request I5R-01-2 Discuss in the relief request how the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Response

Based on the configuration of the nozzle as discussed in Reference 1, the required exam is extremely difficult to perform, and may not yield information that could be used to ascertain the structural integrity of the component. The nozzle design contains reflectors that do not allow a meaningful exam to be performed. Additionally, the performance of this exam would result in significant radiation dose to plant workers. Without the compensating guarantee of obtaining meaningful results, the performance of this exam would be contrary to industry and NRC practices related to maintaining radioactive dose as low as reasonably achievable (ALARA), and represent hardship without a compensating increase in quality and safety.

NRC Request I5R-01-3 Provide a technical basis as to why the ultrasonic testing is not required.

Response

Due to the design of the component, an ultrasonic test does not provide any data that could be used for determining the condition of the nozzle. The complex cladding/socket configuration impedes the ability to obtain ultrasonic test results that would be meaningful or useful in determining the structural integrity and leak tightness of the Standby Liquid Control nozzle.

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ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request I5R-02-1 Do the augmented inspection programs for intergranular stress corrosion cracking Category B-G (Generic Letter 88-01), service water integrity (Generic Letter 89-13), flow accelerated corrosion (Generic Letter 89-08), and high-energy line break (NRC Branch Technical Position MEB 3-1), as described in the initial QCNPS risk-informed ISI submittal dated November 30, 2000 (ADAMS Accession No. ML003776493), remain unaffected by the risk-informed ISI program developed for the fifth interval?

Response

For the QCNPS fifth inservice inspection (ISI) interval, the treatment of the above referenced augmented inspection programs remains unaffected in comparison with the Risk-Informed Inservice Inspection (RISI) Program implemented in the fourth ISI interval. No changes are being made in the fifth ISI interval that alter the implementation methodology of the augmented programs or RISI Program.

NRC Request I5R-02-2 Are the inspection locations in the QCNPS risk-informed ISI programs that have been developed for the fifth 10-year interval the same locations as those in the fourth interval risk-informed ISI programs approved in the NRC staff's January 28, 2004, safety evaluation? If not, please summarize the changes to the program and what caused those changes.

Response

The RISI Program is required to and has been maintained as a living program assessing component and configuration changes and major Probabilistic Risk Assessment (PRA) model revisions throughout the fourth ISI interval. As part of the fifth ISI interval update process, the consequence and degradation assignments and resultant component risk rankings have been confirmed or updated, element selections have been adjusted, and the risk impact assessment has been revised. The final RISI evaluation for the previous fourth ISI interval was Revision 6 dated April 2010. The latest evaluation, Revision 7 dated December 2012, is the current evaluation developed as part of the new fifth interval RISI Program. The changes in inspection locations from the initial fourth interval RISI Program (i.e., Revision 3 dated November 2003) to the new fifth interval RISI Program are summarized in the following tables.

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ATTACHMENT 1 Response to NRC Request for Additional Information QCNPS Unit 1 Selection Summary Fourth Interval Fifth Interval Risk Exams Exams Items Affecting Changes Rank (RISI Rev. 3) (RISI Rev. 7)

High 52 51 Limited Exam Coverage Plant/Component Modifications PRA Model Revisions1 Medium 51 55 Limited Exam Coverage Plant/Component Modifications PRA Model Revisions1 Total 103 106 1

Latest incorporated revision is PRA Model QC110A QCNPS Unit 2 Selection Summary Interval 4 Interval 5 Risk Exams Exams Items Affecting Changes Rank (RISI Rev. 3) (RISI Rev. 7)

High 51 51 Limited Exam Coverage Plant/Component Modifications PRA Model Revisions1 Medium 47 52 Limited Exam Coverage Plant/Component Modifications PRA Model Revisions1 Total 98 103 1

Latest incorporated revision is PRA Model QC110A Limited Exam Coverage - The welds selected for examination were changed in some cases to optimize examination code coverage.

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

PRA Model Revisions - The QCNPS PRA Model applicable to the initial fifth interval RISI Program was revised in January 2011 and issued as Model QC110A.

As the model is updated throughout the interval, impact on the RISI Program is assessed and the program is updated as necessary.

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ATTACHMENT 1 Response to NRC Request for Additional Information In Reference 2, the changes in risk from the pre-RISI Section XI program to the fifth interval RISI Program were provided to demonstrate that the acceptance criteria for delta-core damage frequency (delta-CDF) and delta-large early release frequency (delta-LERF) described in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," were met. As stated above, the latest evaluation developed as part of the new fifth interval RISI Program is Revision 7 dated December 2012. The latest evaluation resulted in slight changes to the delta-CDF and delta-LERF values provided in Reference 2. The revised values, based on the latest evaluation, are listed in the following table. The revised values remain within the acceptance criteria described in Regulatory Guide 1.174.

Change in Risk from Pre-RISI Section XI Program to Fifth Interval RISI Program Station Delta-CDF Delta-LERF QCNPS Unit 1 6.51E-10 1.51E-09 QCNPS Unit 2 1.74E-09 2.08E-09 NRC Request I5R-02-3 If there are changes in the inspection locations for the QCNPS fifth 10-year interval risk-informed ISI programs please provide information for the fifth interval program regarding:

examinations, system, components, degradation mechanisms, class, etc. The information expected should be similar to what is in Tables 2, 3, 4, 5 and 6 of the original submittal of the risk-informed ISI program for the QCNPS third 10-year inservice inspection interval dated November 30, 2000.

Response

A summary of the changes to the inspection locations between the original RISI Program implemented in the fourth ISI interval and the revised program prepared for the fifth ISI interval is contained in the response to NRC Request I5R-02-2. Updated Tables 2 through 6, similar to those provided in the original submittal of the RISI Program, are provided in Attachment 2.

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ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request I5R-02-4 The description and disposition of the gap identified in Attachment 1 of relief request I5R-02 for supporting requirement DA-C12 does not address the requirements in the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008: Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications." Provide verification that a review was conducted that shows support system dependencies are properly captured as required by the standard. In addition, specify whether peer-review or self-assessment results indicate whether this supporting requirement is categorized as capability Category I or not-met. If categorized as "not-met", provide the disposition that ascertains the supporting requirement is now capability Category I.

Response

Supporting requirement (SR) DA-C12 in the American Society of Mechanical Engineers/American Nuclear Society RA-Sa-2009 (i.e., ASME/ANS 2009 PRA Standard) addresses the unavailability of frontline systems due to the unavailability of support systems.

The QCNPS gap listed in Attachment 1 of the submittal was mislabeled as SR DA-C12 rather than SR DA-C13; however, the description and disposition in Attachment 1 are appropriate for SR-DA-C13 of the 2009 Standard. The QCNPS PRA meets the requirements of SR DA-C12 of the 2009 Standard.

In reviewing the self-assessment for the QCNPS Full Power Internal Events (FPIE) PRA, it was determined that a clearer presentation of the self-assessment results should be provided. The self-assessment was reviewed, and corrections and improvements were made to the table of gaps provided in Attachment 1 to the I5R-02 relief request. These changes stem from industry experience on assessing the technical adequacy of PRAs relative to the ASME/ANS 2009 PRA Standard supporting requirements. As a result of this effort, the two gaps originally presented in to relief request I5R-02 are considered met for purposes of RISI, but are replaced by two different gaps. The two former gaps and the current gaps are presented in Table 4-1 below. The gaps originally presented are SC-A5 and DA-C12. Both of these gaps are met with Capability Category (CC) I.

Currently, there are two gaps that are in need of justification to ensure that the PRA is adequate for use in the RISI application. These two gaps affect the model and do not involve issues of documentation. These gaps are SRs DA-C6 and DA-C10. These gaps to the ASME/ANS 2009 PRA Standard have been reviewed to determine which, if any, would merit RISI specific sensitivity studies in the presentation of the application results. The result of this assessment concludes that no additional sensitivity studies are merited, and the gaps do not affect the insights or conclusions of the RISI analysis as represented by relief request I5R-02.

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ATTACHMENT 1 Response to NRC Request for Additional Information Table 4-1: Summary of Previous and Currently Applicable Gaps to RISI ASME/ANS QCNPS CC(1) Disposition Gap SR Document SC-A5 CC I Relief Request None Not a I5R-02, gap for Attachment 1 RISI DA-C13(2) CC I Relief Request None Not a I5R-02, gap for Attachment 1 RISI DA-C6 Gap Revised The number of demands is based on actual Yes Technical operating experience (e.g., Maintenance Adequacy Rule data) if available for a specific system.

Assessment(3) If actual operating experience was not available, then the number of demands was estimated by the system manager (e.g.,

based on surveillance or maintenance test procedures). The PRA Model is judged to appropriately estimate the number of demands for standby equipment for calculating the failure probabilities which will be acceptable for RISI assessments.

DA-C10 Gap Revised The surveillance test procedures are judged Yes Technical to address the appropriate failure modes Adequacy with respect to the estimated number of Assessment(3) demands. No impact on the RISI assessment.

Notes:

• ASME/ANS 2009 Capability Category.

• This SR was mislabeled as DA-C12 in the original submittal.

• Document revised in support of NRC Request I5R-02-4.

NRC Request I5R-06-1 Confirm whether NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," will be used for the inspection of feedwater sparger tee welds and feedwater sparger piping brackets.

Response

Yes, NUREG-0619 will be used for the inspection of feedwater sparger tee welds and feedwater sparger piping brackets.

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ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request I5R-06-2 Section 4.1 item 5 of the BWRVIP-100-A report, "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds," states that fracture toughness values of stainless steel materials that are exposed to a neutron fluence value greater than 1 X 1021 n/cm2 (E > 1 MeV) are lower than those used in Appendix C of the BWRVIP-76 report, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines." Identify whether the core shroud welds and base materials will be exposed to a neutron fluence value greater than 1 X 1021 n/cm2 (E > 1 MeV) during the current ISI interval. Since the inspection frequency in the BWRVIP-76 report is based on fracture toughness values which are not consistent with the BWRVIP-100-A report, the NRC staff requests that the licensee address the following issue:

The inspection frequency and strategy that are specified in Section 3 of the BWRVIP-76 report require further evaluation taking into account the lower fracture toughness values that are specified in the BWRVIP-100-A report.

Response

Not applicable at this time as no shroud or reactor pressure vessel structures are predicted to be exposed to a neutron fluence of greater than 1 X 1021 n/cm2 (E > 1 MeV) during the current 120-month inservice inspection (ISI) interval. Moreover, no current estimations predict crossing this threshold at any point during the period of extended operation.

NRC Request I5R-06-3 Dresden and Quad Cities (D/QC) Safety Evaluation Report, "NUREG-1796, Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2," license renewal application (LRA) commitment #9 in Appendix A of NUREG-1796 states that the licensee should implement the staff approved aging management program for the steam dryers at the D/QC units. In July 2009, the BWRVIP issued a staff approved topical report BWRVIP-139-A, "BWR Vessel and Internals Project, Steam Dryer Inspection and Flaw Evaluation Guidelines." The NRC staff requests that the licensee confirm that it will comply with the guidelines addressed in the BWRVIP-139-A report as per LRA commitment #9 in NUREG-1796.

Response

QCNPS will comply with the guidelines addressed in the BWRVIP-139-A, as discussed in QCNPS UFSAR, Appendix A, Section A.2.10, "Periodic Inspection of Steam Dryers."

NRC Request I5R-06-4 Consistent with the LRA commitment #9, with respect to the aging management program related to the top guide, the licensee should confirm that it will comply with the inspection guidelines addressed in the BWRVIP-26-A and BWRVIP-183 reports.

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ATTACHMENT 1 Response to NRC Request for Additional Information

Response

QCNPS has implemented the guidance provided in BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines," and BWRVIP-183, "Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines." These documents have been incorporated in the QCNPS In-Vessel Visual Inspection (IVVI) program and plant procedures.

References

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Supplement to Quad Cities Nuclear Power Station Fifth Inservice Inspection Interval Relief Request I5R-01," dated November 28, 2012

2. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Supplement to Dresden Nuclear Power Station and Quad Cities Nuclear Power Station Fifth Inservice Inspection Interval Relief Request I5R-02," dated November 28, 2012 Page 8

Attachment 2 Quad Cities Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables RISI Final Report Table 2: Failure Potential Assessment Summary for Unit 1 and Unit 24 Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive System TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC CRD1 ECCS2 X X X X FW X X X HPCI / X RCIC MS RCS3 X RWCU SBLC NOTES:

1. Includes scram discharge volume.

2. Includes core spray (CS), shutdown cooling (SDC), and residual heat removal (RHR).

3. Includes reactor recirculation (RR), reactor pressure vessel (RPV), and jet pump instrument nozzles (JPI).

4. This table shows the assessed failure mechanisms for each system. The RISI program addresses the cumulative impact of all mechanisms that were identified in each system.

TASCS - thermal stratification, cycling and striping, TT - thermal transients, IGSCC - intergranular stress corrosion cracking, TGSCC - transgranular stress corrosion cracking, ECSCC - external chloride stress corrosion cracking, PWSCC - primary water stress corrosion cracking, MIC - microbiologically influenced corrosion, PIT - pitting, CC - crevice corrosion, E-C - erosion-cavitation, FAC - flow accelerated corrosion Page 1

Attachment 2 Quad Cities Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables RISI Final Report Table 3: Number of Elements (Welds) by Risk Category for Unit 1 High Risk Medium Risk Low Risk TOTAL System Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 or 7 All Categories CRD1 53 53 2

ECCS 181 151 61 190 583 FW 53 10 63 HPCI/RCIC 45 105 150 MS 146 8 154 RCS3 101 38 7 69 215 RWCU 26 36 62 SBLC 4 4 TOTAL 53 282 10 406 68 465 1284 NOTES:

1. Includes scram discharge volume.

2. Includes core spray (CS), shutdown cooling (SDC), and residual heat removal (RHR).

3. Includes reactor recirculation (RR), reactor pressure vessel (RPV), and jet pump instrument nozzles (JPI).

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Attachment 2 Quad Cities Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables RISI Final Report Table 4: Number of Elements (Welds) by Risk Category for Unit 2 High Risk Medium Risk Low Risk TOTAL System Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 or 7 All Categories CRD1 53 53 ECCS2 177 145 68 188 578 FW 51 9 60 HPCI/RCIC 27 128 155 MS 142 8 150 RCS3 105 39 7 64 215 RWCU 28 39 67 SBLC 4 4 TOTAL 51 282 9 381 75 484 1282 NOTES:

1. Includes scram discharge volume.

2. Includes core spray (CS), shutdown cooling (SDC), and residual heat removal (RHR).

3. Includes reactor recirculation (RR), reactor pressure vessel (RPV), and jet pump instrument nozzles (JPI).

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Attachment 2 Quad Cities Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables RISI Final Report Table 5: Number of Inspections by Risk Category for Unit 14,5 High Risk Medium Risk Low Risk All Risk Categories Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 or 7 System Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI CRD1 5 0 5 0 ECCS2 29 45 14 19 12 7 18 0 73 71 FW 10 5 0 1 10 6 HPCI / 4 5 11 0 15 5 RCIC MS 74 15 74 15 RCS3 19 5 19 0 38 5 RWCU 8 4 8 0 16 4 SBLC 2 0 2 0 TOTAL 10 5 29 45 0 1 119 48 12 7 63 0 233 106 NOTES:

1. Includes scram discharge volume.

2. Includes core spray (CS), shutdown cooling (SDC), and residual heat removal (RHR).

3. Includes reactor recirculation (RR), reactor pressure vessel (RPV), and jet pump instrument nozzles (JPI).

4. This table provides a comparison of the RISI element selection to the previous Third Intervals 1989 ASME Section XI program (Pre-RISI).

5. This table includes the number of welds previously selected for ASME Section XI (Pre-RISI), but excludes the number of welds previously selected for ASME Section XI (Pre-RISI) that now default to the augmented programs for IGSCC and FAC.

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Attachment 2 Quad Cities Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables RISI Final Report Table 6: Number of Inspections by Risk Category for Unit 24,5 High Risk Medium Risk Low Risk All Risk Categories Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 or 7 System Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI Pre-RISI RISI CRD1 5 0 5 0 ECCS2 28 45 14 18 8 7 16 0 66 70 FW 8 5 0 1 8 6 HPCI / 4 3 10 0 14 3 RCIC MS 73 15 1 0 74 15 RCS3 18 5 19 0 37 5 RWCU 8 4 9 0 17 4 SBLC 2 0 2 0 TOTAL 8 5 28 45 0 1 117 45 8 7 62 0 223 103 NOTES:

1. Includes scram discharge volume.

2. Includes core spray (CS), shutdown cooling (SDC), and residual heat removal (RHR).

3. Includes reactor recirculation (RR), reactor pressure vessel (RPV), and jet pump instrument nozzles (JPI).

4. This table provides a comparison of the RISI element selection to the previous Third Intervals 1989 ASME Section XI program (Pre-RISI).

5. This table includes the number of welds previously selected for ASME Section XI (Pre-RISI), but excludes the number of welds previously selected for ASME Section XI (Pre-RISI) that now default to the augmented programs for IGSCC and FAC.

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