RS-09-081, Additional Information Supporting Risk-Informed Inservice Inspection Relief Request
| ML091900327 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 07/08/2009 |
| From: | Simpson P Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-09-081, TAC ME0225, TAC ME0226 | |
| Download: ML091900327 (10) | |
Text
Exelon Nuclear www.exeloncorp .corn Exel(5n.
4300 Winfield Road Nuclear Warrenville, IL 60555 RS-09-081 July 8, 2009 U . S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457
Subject:
Additional Information Supporting Risk-Informed Inservice Inspection Relief Request
References:
1 . Letter from P. R. Simpson (Exelon Generation Company, LLC) to U . S. NRC, "Third 10-Year Inservice Inspection Interval, Relief Request 13R-01, 'Request for Relief for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10 CFR 50.55a(a)(3)(i),'" dated December 10, 2008
- 2. Letter from M. J . David (U . S. NRC) to C. G. Pardee, "Braidwood Station, Units 1 and 2 - Request for Additional Information Related to Risk-Informed Relief Request 13R-01 (TAC Nos. ME0225 and ME0226)," dated May 11, 2009 In Reference 1, Exelon Generation Company, LLC (EGC) requested authorization to use a risk-informed inservice inspection program as an alternative to the examination program of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, 2001 Edition through the 2003 Addenda, for certain pressure retaining piping welds. The NRC requested additional information to complete review of the relief request in Reference 2. In response to this request, EGC is providing the attached information.
July 8, 2009 U . S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained in this letter . Should you have any questions related to this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803 .
atri Manager - Licensing
Attachment:
Response to Request for Additional Information cc: NRC Senior Resident Inspector NRC Regional Administrator, Region III
ATTACHMENT Response to Request for Additional Information NRC Request 1 The relief request includes the following paragraph :
"The Risk Impact Assessment completed as part of the original baseline RI-ISI Program was an implementation/transition check on the initial impact of converting from a traditional ASME Section XI program to the new RI-ISI methodology. For the Third Interval ISI update, there is no transition occurring between two different methodologies, but rather, the currently approved RI-ISI methodology and evaluation will be maintained for the new interval ."
The NRC staff does not agree with the implication that, if there is no change in methodology, the change in risk assessment is not part of the living process. Furthermore, the submittal is requesting relief to implement a RI-ISI program instead of an ASME program for the third interval, so there is a change from the methodology that would normally be used (i .e., without a relief request) . Regulatory Guide 1 .178, Standard Review Plan 3.9.8, and Electrical Power Research Institute Topical Report 112657 (Refs. 1, 2, and 3) require an evaluation of the change in risk arising from the proposed change in the ISI program . Please provide an estimate of the potential change in risk between the RI-ISI program proposed for implementation in the third interval and the ASME Section XI requirements, which existed prior to the implementation of the first RI-ISI program .
Response
As part of updating the risk informed inservice inspection (RI-ISI) analysis for the third inservice inspection (ISI) interval, the risk impact assessment was updated to confirm the change in risk was maintained within the acceptance guidelines. The previous methodology of the calculation was not changed, and the change in risk was simply re-assessed using the 1989 Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, program prior to RI-ISI and the latest element selection for the third interval RI-ISI program . This same process has been maintained for each revision of the Braidwood RI-ISI report that has been performed to date .
Based on the new interval assessment, the change in risk for Braidwood Unit 1 was 9.64E-08 for core damage frequency (delta-CDF) and 2 .07E-09 for large early release frequency (delta-LERF). For Braidwood Unit 2, the values were 7 .99E-08 for delta-CDF and 1 .52E-09 for delta-LERF . These values are within 1 .00E-06 and 1 .00E-07 acceptance criteria for delta-CDF and delta-LERF, respectively . The change in risk analysis was likewise done at a system level, and no system acceptance criteria are exceeded in the current program using the latest RI-ISI element selections .
NRC Req uest 2 The relief request states :
"As an added measure of assurance, any new systems, portions of systems, or components being included in the RI-ISI program for the Third Inspection Interval will be added to the Risk Impact Assessment performed during the previous interval . These components will be Page 1 of 8
ATTACHMENT Response to Request for Additional Information addressed within the evaluation at the start of the new interval to assure that the new Third Inspection Interval RI-ISI element selection provides an acceptable overall change-in-risk . . ."
The evaluations described above should have already been performed and the results of the evaluations incorporated into the analysis that forms the basis of the relief request. Please provide a brief description of these evaluations and an overview of the results.
Response
For the third ISI interval, the overall scope of the program is similar to the second interval as authorized in Reference 1 . No new systems were added, and no changes to the application of the evaluation methodology were made, that affect the program scoping process. However, the RI-ISI program is required to be maintained, and has been maintained, as a living program assessing component and configuration changes and major probabilistic risk assessment (PRA) model revisions . As part of the third interval update process, the consequence and degradation assignments have been reassessed, component risk rankings have been confirmed or updated, element selections have been adjusted, and the risk impact assessment has been revised . A summary of the element selection results is contained in the response to NRC Request 3 below, and the risk impact results are provided in the response to NRC Request 1 above .
NRC Request 3 The relief request states :
"These portions of the RI-ISI Program have been and will continue to be reevaluated and revised as major revisions of the site PRA occur and modifications to plant configurations are made . . . . The Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, and Element Selection steps encompass the complete living program process . . ."
Please provide the date of the last reevaluation and revision that is described above and a brief description of the results of the reevaluations and revisions undertaken at that date . In addition, please explain how modifications to the units since the date of the last reevaluation and revision may impact the RI-ISI program .
Response
As discussed above in the response to NRC Request 2, the RI-ISI program has been maintained a living program . During the third ISI interval update, a new PRA model revision was incorporated into the program . As a result, some risk rankings were updated and element selections were modified accordingly. In addition, coverage limitations and plant modifications from the second ISI interval were considered in the element selection review process .
The final RI-ISI evaluation of the previous second ISI interval was Revision 4 dated September 2004 . The latest evaluation is the current evaluation developed as part of the new third interval RI-ISI program (i .e., Revision 5 dated December 2008) . No updates have been made since then . The changes in inspection locations since the inception of the RI-ISI program during the second interval to the third interval are summarized below.
ATTACHMENT Response to Request for Additional Information BRAIDWOOD UNIT 1 SELECTION
SUMMARY
Interval 2 Interval 3 Exams Exams Risk (RI-ISI (RI-ISI Rank Rev. 0) Rev. 5) Items Affecting Changes High 81 141 " Limited Exam Coverage
" Plant/Component Modifications
" PRA Model Revisions'
" New Scope due to ASME Code
" Addition of Break Exclusion Region Scope Medium 118 160 " Limited Exam Coverage
" Plant/Component Modifications
" PRA Model Revisions'
" New Scope due to ASME Code
" Addition of Break Exclusion Region Scope Total 199 301
' Latest incorporated revision is PRA Model 6C dated May 2008 BRAIDWOOD UNIT 2 SELECTION SUM MARY Interval 2 Interval 3 Exams Exams Risk (RI-ISI (RI-ISI Rank Rev. 0) Rev. 5) Items Affecting Changes High 109 138 " Limited Exam Coverage
" Plant/Component Modifications
" PRA Model Revisions2
" New Scope due to ASME Code
" Addition of Break Exclusion Region Scope Medium 119 160 " Limited Exam Coverage
" Plant/Component Modifications
" PRA Model Revisions2
" New Scope due to ASME Code
" Addition of Break Exclusion Region Scope Total 228 298 2 Latest incorporated revision is PRA Model 6C dated May 2008 Limited Exam Coverage - The welds selected for examination were changed in some cases to optimize examination code coverage .
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ATTACHMENT Response to Request for Additional Information Plant Modifications - Various plant modifications were installed for both units throughout the interval . These modifications were evaluated for impact to the RI-ISI program, and when applicable, changes to the RI-ISI scope and element selections were made .
PRA Model Revisions - The Braidwood PRA model has been revised since the prior revision of the Braidwood RI-ISI program . Model revision 6C was issued in May 2008, and is the latest revision used in the analysis underpinning the proposed revision to the RI-ISI program .
Significant changes to the model include:
" Changes to Auxiliary Feedwater (AF) system success criteria, based on revised best estimate analyses ;
" Changes to Auxiliary Building flood isolation capability; and
" Data update (e .g., initiating event, failure, and maintenance data) based on recent plant experience .
These changes, and others, contributed to changing the risk profile of the plant such that the following general changes took place in the pipe break consequence assessments:
" Decrease in conditional core damage probability (CCDP) and conditional large early release probability (CLERP) of medium break loss-of-coolant accidents, and
" Decrease in the CCDP and CLERP of steam line breaks inside and outside containment.
New Scope due to ASME Code - For ASME Class 2 components, the IWC-1220 exemption criteria was revised in the code edition applicable to the third ISI interval (i.e., 2001 Edition through the 2003 Addenda) . The third interval ASME Code requires the examination of smaller size piping in the Auxiliary Feedwater (AF) system . As such, the Braidwood RI-ISI program scope was revised to include this new piping . The applicable AF lines were evaluated for degradation and consequence, and risk rankings were assigned. Based on the risk ranking, element selections were made for the high and medium categories .
Addition of Break Exclusion Region Scope - In Revision 4 of the RI-ISI evaluation, the augmented break exclusion region inspection program was added to the RI-ISI program scope.
Evaluation, ranking, and selection requirements were made in accordance with Electric Power Research Institute (EPRI) Technical Report 1006937 (i .e., Reference 2) and the associated NRC safety evaluation (i .e ., Reference 3) .
NRC Rectuest 4 Please explain any changes that were made to the highest estimates in the high category and, if applicable, specify medium consequence segments that moved to the high category .
Response
During a conference call between Exelon Generation Company, LLC (EGC) and the NRC on May 20, 2009, the NRC clarified this request. Specifically, the NRC explained that the intent of this request was to summarize the changes between high and medium risk rank categories .
ATTACHMENT Response to Request for Additional Information This information is provided in the responses to NRC Request 3 above and NRC Request 7 below.
NRCReauest.5 Because the RI-ISI program is a living program, please describe substantive changes to the current RI-ISI program from that submitted in the relief request for the second ISI interval, dated October 16, 2000 (ADAMS Accession No . ML003761986), as a result of industry and plant operating experience . Items discussed in Sections 4.0 and 5.0 of Attachment 1 of the December 10, 2008, relief request do not need to be reiterated .
Response
Changes made to the Braidwood RI-ISI program pertinent to the third interval update are summarized in the response to NRC Request 3 above . No additional substantive changes since the preparation of Revision 5 to the RI-ISI evaluation have been made .
NRC Request 6 Table IWB-2500-1 of ASME Code Section XI, 2001 Edition with 2003 Addenda requires volumetric and/or surface examination of all Category B-F or B-J Pressure Retaining Dissimilar Metal Welds greater than NPS 1 . Based on findings of primary water stress-corrosion cracking in Alloy 82/182 dissimilar metal welds, please provide the following information on inspection plans for these welds .
- a. How many Alloy 82/182 welds are selected for examination?
- b. What examination method(s) are being employed?
- c. What is the frequency of examination?
- d. What is the disposition of limited (<90 percent) examinations?
Response
- a. For Braidwood Station there are currently 28 Alloy 82/182 welds (i.e., 14 per unit) subject to ASME Code,Section XI, Category B-F or B-J inspection requirements . The population of these welds for each unit consists of four reactor pressure vessel (RPV) nozzle-to-safe end hot leg welds, four RPV nozzle-to-safe end cold leg welds, and six pressurizer safe end-to-nozzle welds mitigated through application of full structural weld overlays.
- b. The RPV nozzle-to-safe end hot and cold leg dissimilar metal welds have been inspected, and until mitigation is completed, will continue to be inspected by bare metal visual inspection in accordance with Reference 4 and ultrasonic (UT) methods demonstrated through the EPRI Performance Demonstration Initiative (PDI) Program in accordance with Reference 5.
- c. The frequency of examinations is as follows .
ATTACHMENT Response to Request for Additional Information Braidwood Station Alloy 82/182 Weld Inspectio n Frequency Bare Visual Volumetric Governing Weld Type Notes Examination Examination Require ment RPV Hot Leg Every Five N-722 (Visual) Eddy current also Every Outage Nozzle-to-Safe Years Until performed on inside Until Mitigated MRP-139 (UT)
End Mitigated diameter surface RPV Cold Leg Once Per Every Six N-722 (Visual) Eddy current also Nozzle-to-Safe Interval Until Years Until inside End Mitigated Mitigated MRP-139 (UT) diameter dsurface Pressurizer Examine All ASME XI Not Required Nozzle-to-Safe Within Next Nonmandatory None (Mitigated)
End Overlays Two Outages Appendix Q
- d. The examination volume coverage achieved for the 28 alloy 82/182 welds (post weld overlay examinations for welds associated with pressurizer nozzle-to-safe end welds) was greater than 90%.
NRC Request 7 Please provide a list of the specific welds that are selected for examination in the third ISI interval .
- a. If the welds selected for examination are different than those selected for the second ISI interval, please describe the reasons for the selections.
Response
During a conference call between EGC and the NRC on May 20, 2009, the NRC clarified this request. Specifically, the NRC explained that the intent of this question was to summarize the changes between the risk rank categories subject to inspection under the EPRI RI-ISI methodology.
A summary of the changes to inspection locations between the original RI-ISI program implemented in the second interval and the revised program prepared for the third interval is contained in the response to NRC Request 3 above. The following tables provide selection comparisons for both units .
ATTACHMENT Response to Request for Additional Information BRAIDWOOD UNIT 1 SELECTION COMPARISON Minimum RI-ISI Elements RI-ISI Elements Section XI RI-ISI Selected for Selected for Welds Selected Risk Elements to Element Interval 3 Interval 2 for Interval 2 Category Select for Populations Interval 3 (RI-ISI Rev. 5) (RI-ISI Rev. 0 ) (Pre-RI-ISI) 1 128 32.0 32 0 49 2 218 54.5 109 49 145 3 0 0 .0 0 32 0 4 1144 114 .4 118 85 435 5 418 41 .8 42 33 77 TOTALS : 1908 242.7 301 199 706 BRAIDWOO D U NIT 2 S ELECTION COM PARISON Minimum RI-ISI Elements RI-ISI Elements Section XI RI-ISI Selected for Selected for Welds Selected Risk Elements to Element Interval 3 Interval 2 for Interval 2 Category Select for Populations Interval 3 (RI-ISI Rev. 5) (RI-ISI Rev. 0) (Pre-RI-ISI) 1 236 59.0 59 0 37 2 218 54.5 79 50 111 3 0 0 0 59 0 4 1154 115 .4 119 89 429 5 404 40 .4 41 30 118 L
- - 228 TOTALS : - 2012 l 269 .3 - ' - 298 695 NRC Request 8 Please provide a list of any augmented inspection programs which have been subsumed into the RI-ISI program or added since the second ISI interval RI-ISI program was instituted.
Response
One augmented inspection program has been subsumed into the RI-ISI program since it was initially implemented. The Break Exclusion Region augmented inspection program for welds in the main steam and feedwater systems was subsumed into the second interval RI-ISI program using the EPRI methodology developed under Reference 2.
No other changes as to how the original RI-ISI program interfaces with other station augmented inspection programs have been made . Other augmented programs continue to be maintained outside of the RI-ISI program as the EPRI guidance does not provide rules to subsume all augmented inspection programs .
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ATTACHMENT Response to Request for Additional Information References
- 1. Letter from A. J . Mendiola (NRC) to O . D . Kingsley (Exelon Generation Company, LLC),
"Braidwood Station, Units 1 and 2 - Interval 2 Inservice Inspection Program - Relief Request 12R-39, Alternative to the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code Section XI Requirements for Class 1 and Class 2 Piping Welds (TAC Nos. MB0506 and MB0507)," dated February 20, 2002
- 2. Electric Power Research Institute Technical Report 1006937, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs, Rev. 0-A," dated August 2002
- 3. Letter from C. F. Holden (NRC) to G. L. Vine (Electric Power Research Institute), "Safety Evaluation of Topical Report TR-1006937, 'Extension of the EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs' (TAC No. MB1344)," dated June 27, 2002
- 4. American Society of Mechanical Engineers Code Case N-722, "Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials," dated July 5, 2005