RS-13-107, Supplement to Fifth Inservice Inspection Interval Relief Requests I5R-01, I5R-02, and I5R-07
| ML13154A248 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/03/2013 |
| From: | Simpson P Exelon Generation Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RS-13-107 | |
| Download: ML13154A248 (14) | |
Text
4301 0JU RS-1 3-107 June 3, 2013 10 CFR 50.55a U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C. 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249
Subject:
Supplement to Dresden Nuclear Power Station Fifth Inservice Inspection Interval Relief Requests 15R-01, 15R-02, and 15R-07
References:
1.
Letter from D. M. Gullott (Exelon Generation Company, LLC (EGC)) to U. S.
NRC, "Dresden Nuclear Power Station, Units 2 and 3, Fifth Interval Inservice Inspection Program Plan and Relief Requests," dated September 28, 2012 2.
Letter from P. R. Simpson (EGC) to U. S. NRC, "Supplement to the Dresden Nuclear Power Station Fifth Inservice Inspection Interval Relief Request 15R-01," dated November 19, 2012 3.
Letter from P. R. Simpson (EGC) to U. S. NRC, "Supplement to the Dresden Nuclear Power Station and Quad Cities Nuclear Power Station Fifth Inservice Inspection Interval Relief Request 15R-02," dated November 28, 2012 In the referenced letters, EGC submitted relief requests and supplemental information associated with the fifth inservice inspection (ISI) interval for Dresden Nuclear Power Station (DNPS), Units 2 and 3. During the NRC's review of the referenced documents, the NRC found that additional information is required to support its review of Relief Requests 15R-01, 15R-02, and 15R-07. The requested information is provided in the attachments to this letter.
j 1
RS-13-107 10 CFR SO.SSa June 3,2013 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D.C. 20SSS-0001
Subject:
Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-2S NRC Docket Nos. SO-237 and S0-249 Supplement to Dresden Nuclear Power Station Fifth Inservice Inspection Interval Relief Requests ISR-01, ISR-02, and ISR-07
References:
- 1.
Letter from D. M. Gu"ott (Exelon Generation Company, LLC (EGC>> to U. S.
NRC, "Dresden Nuclear Power Station, Units 2 and 3, Fifth Intervallnservice Inspection Program Plan and Relief Requests," dated September 28,2012
- 2.
Letter from P. R. Simpson (EGC) to U. S. NRC, "Supplement to the Dresden Nuclear Power Station Fifth Inservice Inspection Interval Relief Request ISR-01," dated November 19,2012
- 3.
Letter from P. R. SimpsQn (EGC) to U. S. NRC, "Supplement to the Dresden Nuclear Power Station and Quad Cities Nuclear Power Station Fifth Inservice Inspection Interval Relief Request ISR-02," dated November 28, 2012 In the referenced letters, EGC submitted relief requests and supplemental information associated with the fifth inservice inspection (lSI) interval for Dresden Nuclear Power Station (DNPS), Units 2 and 3. During the NRC's review of the referenced documents, the NRC found that additional information is required to support its review of Relief Requests ISR-01, ISR-02, and ISR-07. The requested information is provided in the attachments to this letter.
June 3, 2013 U. S. Nuclear Regulatory Commission Page 2 EGC continues to request approval of these relief requests by September 28, 2013, to support examination of components during the DNPS, Unit 2 Fall 2013 refueling outage which is currently scheduled to occur within the fifth 10-year ISI interval.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.
Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments:
1.
Response to NRC Request for Additional Information 2.
Dresden Nuclear Power Station Fifth ISI Interval Risk-Informed Inservice Inspection (RISI)
Summary Tables June 3,2013 U. S. Nuclear Regulatory Commission Page 2 EGC continues to request approval of these relief requests by September 28, 2013, to support examination of components during the DNPS, Unit 2 Fall 2013 refueling outage which is currently scheduled to occur within the fifth 10-year lSI interval.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this letter, please contact IVlr. Mitchel A. Mathews at (630) 657-2819.
Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments:
- 1. Response to NRC Request for Additional Information
- 2. Dresden Nuclear Power Station Fifth lSI Interval Risk-Informed Inservice Inspection (RISI)
Summary Tables
ATTACHMENT 1 Response to NRC Request for Additional Information Based on the staff's review of Relief Request 15R-01:
NRC Request 15R-01-1.
Discuss in the Relief Request why the proposed alternative provides reasonable assurance of structural integrity or leak tightness of the subject component(s).
Exelon Generation Company, LLC (EGC) Response:
The proposed alternative looks for and 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 of 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 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 Dresden Nuclear Power Station (DNPS), Units 2 and 3 from this location.
NRC Request 15R-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 quality and safety.
EGC Response:
Based on the configuration of the nozzle as discussed in Reference 2, 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 guarantee of obtaining meaningful results, the performance of this exam would be contrary to industry and NRC practices related to maintaining radiation dose as low as reasonably achievable (ALARA), and represent hardship without a compensating increase in quality and safety.
Page 1 of 7 ATTACHMENT 1 Response to NRC Request for Additional Information Based on the staffs review of Relief Request ISR-01:
NRC Request ISR-01-1.
Discuss in the Relief Request why the proposed alternative provides reasonable assurance of structural integrity or leak tightness of the subject component(s).
Exelon Generation Company. LLC (EGC) Response:
The proposed alternative looks for and 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 of 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 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 Dresden Nuclear Power Station (DNPS), Units 2 and 3 from this location.
NRC Request ISR-01-2.
EGC Response:
Discuss in the Relief Request how the specified requirements of this section would resultin hardship or unusual difficulty without a compensating increase In the level of quality and quality and safety.
Based on the configuration of the nozzle as discussed in Reference 2, 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 guarantee of obtaining meaningful results. the performance of this exam would be contrary to industry and NRC practices related to maintaining radiation dose as low as reasonably achievable (ALARA). and represent hardship without a compensating increase in quality and safety.
Page 1 of7
ATTACHMENT I Response to NRC Request for Additional Information NRC Request 15R-01 -3.
Provide a technical basis as to why the ultrasonic testing is not required.
EGC 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 EGC's 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.
Based on the staff's review of Relief Request 15R-02:
NRC Request 15R-02-1.
Do the augmented inspection programs for IGSCC Category B-G (Generic Letter 88-01), service water integrity (Generic Letter 89-13), FAC (Generic Letter 89-09), and high-energy line break (HELB)(USNRC Branch Technical Position MEB 3 -1) remain unaffected (as described in the initial DNPS RI-lSl submittal dated October 18, 2000) by the RI-ISI program developed for the fifth interval?
EGC Response:
For the DNPS 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 15R-02-2.
Are the inspection locations in the DNPS RI-ISI programs that have been developed for the fifth 10-year interval the same locations as those in the fourth interval RI-ISI programs approved in the NRC staff's September 4, 2003, safety evaluation? If not, please summarize the changes to the program and what caused those changes.
EGC 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 5 dated January 2010. The latest evaluation, Revision 6, dated December 2012, is the current evaluation developed as part of the new fifth interval RISI Program. The changes in inspection Page 2 of 7 ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request 15R-01-3.
Provide a technical basis as to why the ultrasonic testing is not required.
EGC 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 EGC's 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.
Based on the staffs review of Relief Request 15R*02:
NRC Request 15R-02-1.
Do the augmented inspection programs for IGSCC Category B-G (Generic Letter 88-01), service water integrity (Generic Letter 89-13), FAC (Generic Letter 89-09), and high-energy line break (HELB)(USNRC Branch Technical Position MEB 3-1) remain unaffected (as described in the initial DNPS RI-ISI submittal dated October 18,2000) by the RI-ISI program developed for the fifth interval?
EGC Response:
For the DNPS fifth inservice inspection (lSI) 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 lSI interval. No changes are being made in the fifth lSI interval that alter the implementation methodology of the augmented programs or RISI Program.
NRC Request 15R-02-2.
Are the inspection locations in the DNPS RI-ISI programs that have been developed for the fifth 10-year interval the same locations as those in the fourth interval RI-ISI programs approved in the NRC staff's September 4, 2003, safety evaluation? If not, please summarize the changes to the program and what caused those changes.
EGC 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 lSI interval. As part of the fifth lSI 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 R1SI evaluation for the previous fourth lSI interval was Revision 5 dated January 2010. The latest evaluation, Revision 6, dated December 2012. is the current evaluation developed as part of the new fifth interval RISI Program. The changes in inspection Page 2 of7
ATTACHMENT 1 Response to NRC Request for Additional Information locations from the initial fourth interval RISI Program (i.e., Revision 3 dated December 2002) to the new fifth interval RISI Program are summarized in the Tables 1 and 2 below.
Table 1: DNPS Unit 2 Selection Summa R
Intend 4 In al 5
^;
Rart^
tts s
tt i
'04ii (RISIRe to High 49 58 n
Limited Exam Coverage n
Plant/Component Modifications n
PRA Model Revisions' Medium 47 55 n
Limited Exam Coverage n
Plant/Component Modifications n
PRA Model Revisions' Total 96 113 Latest incorporated revision is PRA Model DR209A Table 2: DNPS Unit 3 Selection Summa Risk lnterval4;,
II"nterval Rangy
" ` Exams-Exams-l Affecting Chan RIM N^1)'
Mt Revs.
High 41 50 n
Limited Exam Coverage n
Plant/Component Modifications n
PRA Model Revisions' Medium 50 56 n
Limited Exam Coverage n
Plant/Component Modifications n
PRA Model Revisions' Total 91 106
' Latest incorporated revision is PRA Model DR209A 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.
Page 3 of 7 ATIACHMENT1 Response to NRC Request for Additional Information locations from the initial fourth interval RISI Program (Le., Revision 3 dated December 2002) to the new fifth interval RISI Program are summarized in the Tables 1 and 2 below.
High 49 Medium 47 55 Total 96 113 Limited Exam Coverage Plant/Component Modifications PRA Model Revisions 1 Limited Exam Coverage Plant/Component Modifications PRA Model Revisions 1 1 Latest incorporated revision is PRA Model DR209A Table 2: DNPS Unit 3 Selection Summary High 41 50* Limited Exam Coverage Medium 50 Total 91 56 106 Plant/Component Modifications PRA Model Revisions 1 Limited Exam Coverage Plant/Component Modifications PRA Model Revisions 1 1 Latest incorporated revision is PRA Model DR209A 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 lSI 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.
Page 3 of 7
ATTACHMENT I Response to NRC Request for Additional Information PRA Model Revisions - The DNPS PRA Model applicable to the initial fifth interval RISI Program was revised in December 2009 and issued as Model DR209A. This revision of the model was incorporated into Revision 5 of the RISI Program in January 2010. As the model is updated throughout the interval, impact on the RISI Program is assessed and the program is updated as necessary.
In Reference 3 above, 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 6 dated December 2012. The latest evaluation resulted in slight changes to the delta-CDF and delta-LERF values provided in Reference 3. The revised values, based on the latest evaluation, are listed in Table 3 below. The revised values remain within the acceptance criteria described in Regulatory Guide 1.174.
Table 3: Change in Risk from Pre-RISI Section XI Pro ram to Fifth Interval RISI Proar Station `:
1
)
it-l C Ita-IL R DNPS Unit 2 4.62E-9 2.03E-09 DNPS Unit 3 3.00E-09 9.85E-09 NRC Request 15R-02-3.
If there are changes in the inspection locations for the DNPS fifth 10-year interval RI-ISI programs please provide information for the fifth interval program regarding: examinations
/system/components /degradation mechanisms /class, etc.
similar to that provided in Tables 2, 3, 4, 5 and 6 of the original submittal of the RI-ISI program for the DNPS third 10-year inservice inspection interval dated October 18, 2000 (ADAMS Accession No. ML003762371).
EGC 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 No. 15R-02-2 above. Updated Tables 2 through 6 similar to those provided in the original submittal of the RISI Program are provided in.
Page 4 of 7 ATIACHMENT 1 Response to NRC Request for Additional Information PRA Model Revisions - The DNPS PRA Model applicable to the initial fifth interval RISI Program was revised in December 2009 and issued as Model DR209A. This revision of the model was incorporated into Revision S of the RISI Program in January 2010. As the model is updated throughout the interval, impact on the RISI Program is assessed and the program is updated as necessary.
In Reference 3 above, 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 6 dated December 2012. The latest evaluation resulted in slight changes to the delta-CDF and delta-LERF values provided in Reference 3. The revised values, based on the latest evaluation, are listed in Table 3 below. The revised values remain within the acceptance criteria described in Regulatory Guide 1.174.
DNPS Unit 2 4.62E-9 2.03E-09 DNPS Unit 3 3.00E-09 9.8SE-09 NRC Request 15R-02-3. If there are changes in the inspection locations for the DNPS fifth 10-year interval RI-ISI programs please provide information for the fifth interval program regarding: examinations
/systemlcomponents /degradation mechanisms Iclass, etc.
similar to that provided in Tables 2, 3, 4, 5 and 6 of the original submittal of the RI-/SI program for the DNPS third 10-year inservice inspection interval dated October 18, 2000 (ADAMS Accession No. MLOO3762371).
EGC Response:
A summary of the changes to the inspection locations between the original RISI Program implemented in the fourth lSI interval and the revised program prepared for the fifth lSI interval is contained in the response to NRC Request No. ISR-02-2 above. Updated Tables 2 through 6 similar to those provided in the original submittal of the RISI Program are provided in.
Page 4 of7
ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request 15R-02-4.
The description and disposition of the gap identified in of relief request 15R-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.
EGC Response:
Supporting requirement (SR) DA-C12 in the American Society of Mechanical Engineers/American Nuclear Society RA-Sa-2009 (ASME/ANS 2009 PRA Standard) addresses the unavailability of frontline systems due to the unavailability of support systems. The DNPS gap listed in Attachment 1 of the submittal refers to supporting requirement DA-C12 in the RA-Sb-2005 version of the standard, not the 2009 Standard. DA-C12 in the 2009 Standard is equivalent to DA-CI1 a in the 2005 version of the standard. The DNPS FPIE PRA used in the RISI analysis meets DA-C1 la with a Capability Category I.
Based on the staff's review of Relief Request 15R-07:
NRC Request 15R-07-1.
The staff requests that the licensee identify whether there are any furnace-sensitized stainless steel vessel attachment welds associated with the RVI components at the Dresden Nuclear Power Station, Units 2 and 3. It is requested that the licensee provide an explanation regarding the type of inspection program and any additional augmented inspection program that are implemented for any existing furnace-sensitized stainless steel attachment welds in these boiling water reactor (BWR) units.
EGC Response:
Dresden Nuclear Power Station (DNPS), Unit 2 does have furnace-sensitized stainless steel vessel attachment welds associated with the RVI components, but DNPS, Unit 3 does not. The furnace-sensitized stainless steel vessel attachment welds on Unit 2 will be examined in accordance with the inspection recommendations of BWRVIP-48-A, "Vessel ID Attachment Weld Inspection and Evaluation Guidelines," as discussed in DNPS Updated Final Safety Analysis Report (UFSAR), Appendix A, Section A.1.4, "BWR Vessel ID Attachment Welds."
Page 5 of 7 ATIACHMENT1 Response to NRC Request for Additional Information NRC Request 15R-02-4.
The description and disposition of the gap identified in of relief request 15R*02 for supporting requirement DA*C12 does not address the requirements in the American Society of Mechanical Engineers (ASME)IAmerican Nuclear Society (ANS) RA-Sa-2009, "Addenda to ASMEIANS RA-S-200B:
EGC Response:
Standard for Level11Large 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.
Supporting requirement (SR) DA-C12 in the American Society of Mechanical Engineers/American Nuclear Society RA-Sa-2009 (ASMEIANS 2009 PRA Standard) addresses the unavailability of frontline systems due to the unavailability of support systems. The DNPS gap listed in Attachment 1 of the submittal refers to supporting requirement DA-C12 in the RA-Sb-2005 version of the standard, not the 2009 Standard. DA-C12 in the 2009 Standard is equivalent to DA-C 11 a in the 2005 version of the standard. The DNPS FPIE PRA used in the RISI analysis meets DA-C11 a with a Capability Category I.
Based on the staff's review of Relief Request 15R-07:
NRC Request /5R-07-1.
The staff requests that the licensee identify whether there are any furnace-sensitized stainless steel vessel attachment welds associated with the RVI components at the Dresden Nuclear Power Station, Units 2 and 3. It is requested that the licensee provide an explanation regarding the type of inspection program and any additional augmented inspection program that are implemented for any existing furnace-sensitized stainless steel attachment welds in these boiling water reactor (BWR) units.
EGC Response:
Dresden Nuclear Power Station (DNPS), Unit 2 does have furnace-sensitized stainless steel vessel attachment welds associated with the RVI components, but DNPS, Unit 3 does not. The furnace-sensitized stainless steel vessel attachment welds on Unit 2 will be examined in accordance with the inspection recommendations of BWRVIP-48-A, "Vessel 10 Attachment Weld Inspection and Evaluation Guidelines," as discussed in DNPS Updated Final Safety Analysis Report (UFSAR), Appendix A, Section A 1.4, "BWR Vessel 10 Attachment Welds."
Page 5 of 7
ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request 15R-07-2.
The staff requests that the licensee 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.
EGC Response:
Yes. 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.
NRC Request 15R-07-3.
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 Mel/) 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 I X 1021 n1cm2 (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 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.
EGC 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 ISI interval. Moreover, no current estimations predict crossing this threshold at any point during the future.
Page 6 of 7 ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request /SR-07-2.
The staff requests that the licensee confirm whether NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Retum Line Nozzle Cracking." will be used for the inspection of feedwater sparger tee welds and feedwater sparger piping brackets.
EGC Response:
Yes. 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.
NRC Request ISR-07-3.
Section 4.1 item S 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 nlcm2 (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 nlcm2 (E > 1 MeV) during the current lSI 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 staff requests that the licensee address the following issue:
EGC Response:
The inspection frequency and strategy that are specified in Section 3 of the BWRVlP*76 report require further evaluation taking into account the lower fracture toughness values that are specified in the BWRVIP-100-A report.
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 lSI interval. Moreover, no current estimations predict crossing this threshold at any point during the future.
Page 6 of 7
ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request 15R-07-4.
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 Commitment (LRA) #9 in Appendix A of NUREG-1796 states that the licensee should implement the staff approved AMP 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 staff requests that the licensee confirm that it will comply with the guidelines addressed in the BWRVIP-139-A report as per LRA #9 in NUREG-1796.
EGC Response:
DNPS will comply with the guidelines addressed in the BWRVIP-139-A, "BWR Vessel and Internals Project, Steam Dryer Inspection and Flaw Evaluation Guidelines," as discussed in DNPS UFSAR, Appendix A, Section A.2.10, "Periodic Inspection of Steam Dryers."
NRC Request 15R-07-5.
Consistent with the LRA commitment #9, with respect to the AMP 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.
EGC Response:
DNPS 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," as discussed in DNPS UFSAR, Appendix A, Section A.1.9, "BWR Vessel Internals."
Page 7of7 ATTACHMENT 1 Response to NRC Request for Additional Information NRC Request ISR*07*4.
Dresden and Quad Cities (DIQC) 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 Commitment (LRA) #9 in Appendix A of NUREG-1796 states that the licensee should implement the staff approved AMP for the steam dryers at the DIQC 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 staff requests that the licensee confirm that it will comply with the guidelines addressed in the BWRVIP-139-A report as per LRA #9 in NUREG-1796.
EGC Response:
DNPS will comply with the guidelines addressed in the BWRVIP-139-A, "BWRVessel and Internals Project, Steam Dryer Inspection and Flaw Evaluation Guidelines," as discussed in DNPS UFSAR, Appendix A, Section A.2.10, "Periodic Inspection of Steam Dryers."
NRC Request ISR-07-S.
Consistent with the LRA commitment #9, with respect to the AMP related to the top guide, the licensee should confirm that it will comply with the inspection guidelines addressed in the BWRVIP-26*A and BWRVlp*183 reports.
EGC Response:
DNPS 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," as discussed in DNPS UFSAR, Appendix A, Section A.1.9, "BWR Vessel Internals."
Page 7 of 7
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 1 of 5 RISI Final Report Table 2: Failure Potential Assessment Summary for Unit 2 and Unit 35 h
Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive System CRD' ECCS2 FW HPCI MS3 RCS4 TASCS X
X X
X TT X
X X
X IGSCC X
X X
TGSCC ECSCC PWSCC MIC PIT CC E-C FAC X
X X
1.
Includes scram discharge volume (CRDSD).
2.
Includes CS, LPCI, RHS, and RHSP.
3.
Includes isolation condenser (ISCOCR, ISCOSS).
4.
Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (LVLA, LVLB, UVLA, UVLB).
5.
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 X
X X
X X
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 1 of 5 RISI Final Report Table 2: Failure Potential Assessment Summary for Unit 2 and Unit 35 Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive System TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC CR01 ECCS2 X
X X
FW X
X X
HPCI X
MS3 X
X X
X RCS4 X
X RWCU X
X SBLC X
SOC X
X X
NOTES:
- 1. Includes scram discharge volume (CROSO).
- 3. Includes isolation condenser (ISCOCR, ISCOSS).
- 4. Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (LVLA, LVLB, UVLA, UVLB).
- 5. 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
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 2 of 5 RISI Final Report Table 3: Number of Elements (Welds) by Risk Category for Unit 2 High Risk Medium Risk Low Risk TOTAL System CRD' ECCS2 FW HPCI MS3 RCS4 RWCU SBLC SDC TOTAL Category 1 Category 2 16 115 10 13 47 254 Category 3 Category 4 12 121 95 15 426 Category 5 32 7
6 47 Category 6 or 7
64 435 142 86 87 14 1
24 853 All Categories 64 638 56 186 292 304 50 49 71 1710 3
48 66 5
82 35 2
53 148 11 45 NOTES:
1.
Includes scram discharge volume (CRDSD).
2.
Includes CS, LPCI, RHS, and RHSP.
3.
Includes isolation condenser (ISCOCR, ISCOSS).
4.
Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (LVLA, LVLB, UVLA, UVLB).
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 2 of 5 RISI Final Report Table 3: Number of Elements (Welds) by Risk Category for Unit 2 High Risk Medium Risk Low Risk Category 6 or TOTAL All System Category 1 Category 2 Category 3 Category 4 Category 5 7
Categories CR01 64 ECCS2 53 148 2
435 FW 45 11 HPCI 12 32 142 MS3 3
16 66 121 86 RCS4 115 95 7
87 RWCU 10 5
15 6
14 SBle 13 35 1
SOC 47 24 TOTAL 48 254 82 426 47 853 NOTES:
- 1. Includes scram discharge volume (CROSO).
- 3. Includes isolation condenser (ISCOCR, ISCOSS).
- 4. Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (lVLA, lVlB, UVLA, UVlB).
64 638 56 186 292 304 50 49 71 1710 I
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 3 of 5 RISI Final Report Table 4: Number of Elements (Welds) by Risk Category for Unit 3 High Risk Medium Risk Low Risk TOTAL System CRD' ECCS2 FW HPCI MS3 RCS4 RWCU SBLC SDC TOTAL Category 1 51 Category 2 2
10 17 44 100 Category 3 89 Category 4 9
112 125 15 57 465 Category 5 29 9
6 46 Category 6 or 7
51 484 147 95 101 17 1
19 915 All Categories 51 660 57 187 296 235 42 75 63 1666 5
74 4
2 27 147 11 46 NOTES:
- 1. Includes scram discharge volume (CRDSD).
2.
3.
Includes isolation condenser (ISCOCR, ISCOSS).
4.
Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (LVLA, LVLB, UVLA, UVLB).
ATIACHMENT2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 30f5 RISI Final Report Table 4: Number of Elements (Welds) by Risk Category for Unit 3 High Risk Medium Risk low Risk Category 6 or TOTAL All System Category 1 Category 2 Category 3 Category 4 Category 5 7
Categories CR01 51 ECCS2 27 147 2
9 29 147 MS3 5
10 74 112 95 RCS4 125 9
101 RWCU 4
15 6
17 SBlC 17 57 1
SOC 44 19 TOTAL 51 100 89 465 46 915 NOTES:
- 1. Includes scram discharge volume (CROSO).
- 3. Includes isolation condenser (ISCOCR, ISCOSS).
- 4. Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (lVLA, LVLB, UVLA, UVLB).
51 660 57 187 296 235 42 75 63 1666 i
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 4 of 5 RISI Final Report Table 5: Number of Inspections by Risk Category for Unit 25,6 High Risk Medium Risk Low Risk All Risk Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 or 7 Categories 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 19 14 12 16 39 0
70 30 FW 7
6 1
2 8
8 HPCI 4
4 13 0
19 6
MS3 1
1 3
3 4
14 33 13 19 0
60 31 RCS4 2
2 21 13 4
1 21 0
48 16 RWCU 4
2 3
0 7
2 SBLC 5
4 8
4 13 8
SDC 12 12 5
0 17 TOTAL 8
7 41 35 5
16 82 50 6
5 105 0
247 NOTES:
- 1. Includes scram discharge volume (CRDSD).
2.
Includes CS, LPCI, RHS, and RHSP.
3.
Includes isolation condenser (ISCOCR, ISCOSS).
4.
Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (LVLA, LVLB, UVLA, UVLB).
- 5. This table provides a comparison of the RISI element selection to the previous Third Interval's 1989 ASME Section XI program (Pre-RISI).
- 6. 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.
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page4of5 RISI Final Report Table 5: Number of Inspections by Risk Category for Unit 25,6 High Risk Medium Risk Low Risk All Risk Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 or 7 Categories System Pre-RISI Pre-RISI Pre-RISI Pre-RISI Pre-RISI RISI Pre-RISI RISI RISI RISI RISI RISI CR01 5
0 ECCS2 19 14 12 16 39 0
FW 7
6 1
2 HPCI 4
2 2
4 13 0
MS3 1
1 3
3 4
14 33 13 19 0
RCS4 2
2 21 13 4
1 21 0
RWCU 4
2 3
0 SBlC 5
4 8
4 SOC 12 12 5
0 TOTAL 8
7 41 35 5
16 82 50 6
5 105 0
NOTES:
- 1. Includes scram discharge volume (CRDSO).
- 3. Includes isolation condenser (ISCOCR, ISCOSS).
- 4. Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB). and RPV level instrument nozzles (lVLA, lVlB, UVLA, UVlB).
Pre-RISI 5
70 8
19 60 48 7
13 17 247
- 5. This table provides a comparison of the RISI element selection to the previous Third Interval's 1989 ASME Section XI program (Pre-RISI).
- 6. 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.
RISI 0
30 8
6 31 16 2
8 12 113
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 5 of 5 RISI Final Report Table 6: Number of Inspections by Risk Category for Unit 35,6 High Risk I
Medium Risk Category 1 Category 2 Category 3 1 Category 4 (
Category 5 Low Risk Category 6 or 7 All Risk Categories System RISI I Pre-RISI I RISI RISI RISI RISI Pre-RISI Pre-RISI Pre-RISI Pre-RISI RISI RISI Pre-RISI Pre-RISI CRD' ECCS2 FW 6
11 7
1 2
19 16 4
40 0
23 8
HPCI MS3 1
RCS4 2
1 2
2 3
12 14 3
1 0
18 0
55 0
72 14 15 25 5
31 15 1
8 19 0
16 3
2 6
14 5
10 5
11 5
6 2
11 11 TOTAL 1
9 1
8 27 1
24 1 9
1 18 1 112 1 52 1 5
1 4
1 109 1 0
271 1
106 NOTES:
- 1. Includes scram discharge volume (CRDSD).
2.
Includes CS, LPCI, RHS, and RHSP.
3.
Includes isolation condenser (ISCOCR, ISCOSS).
4.
Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (LVLA, LVLB, UVLA, UVLB).
- 5. This table provides a comparison of the RISI element selection to the previous Third Interval' s 1989 ASME Section XI program (Pre-RISI).
- 6. 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.
ATTACHMENT 2 Dresden Nuclear Power Station Fifth Interval Risk-Informed Inservice Inspection (RISI) Summary Tables Page 5 of 5 RISI Final Report Table 6: Number of Inspections by Risk Category for Unit 35.6 High Risk Medium Risk Low Risk All Risk Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 or 7 Categories System Pre-RISI Pre-RISI Pre-RISI Pre-RISI Pre-RISI RISI Pre-RISI RISI RISI RISI RISI RISI CR01 4
0 ECCS2 11 7
19 16 40 0
FW 8
6 1
2 HPCI 2
2 2
3 14 0
MS3 1
2 1
1 8
16 30 12 15 0
RCS4 44 14 3
1 25 0
RWCU 3
2 5
0 SBLC 5
5 14 6
SOC 10 11 6
0 TOTAL 9
8 27 24 9
18 112 52 5
4 109 0
NOTES:
- 1. Includes scram discharge volume (CROSO).
- 3. Includes isolation condenser (ISCOCR, ISCOSS).
- 4. Includes reactor recirculation (RR), reactor head vent (RHV), jet pump (JPIA, JPIB), and RPV level instrument nozzles (LVLA, LVLB, UVLA, UVLB).
Pre-RISI 4
70 9
18 55 72 8
19 16 271
- 5. This table provides a comparison of the RISI element selection to the previous Third Interval's 1989 ASME Section XI program (Pre-RISI).
- 6. 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.
RISI o '
23 8
5 I
31
- 15 2
11 11 106