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{{Adams
#REDIRECT [[RS-12-040, Response to Request for Additional Information Related to Braidwood Station Relief Request 13R-08]]
| number = ML120730196
| issue date = 03/12/2012
| title = Braidwood Station, Units 1 & 2, Response to Request for Additional Information Related to Braidwood Station Relief Request 13R-08
| author name = Gullott D M
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000456, 05000457
| license number = NPF-072, NPF-077
| contact person =
| case reference number = RS-12-040, TAC ME6024, TAC ME6025
| document type = Letter
| page count = 30
| project = TAC:ME6024, TAC:ME6025
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:Exelon Generation Company, LLC www.exeloncorp.com 4300 Winfield Road Warrenville, IL 60555 Nuclear RS-12-040 March 12, 2012 10 CFR 50.55a U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-001 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:==
Response to Request for Additional Information Related to Braidwood Station Relief Request 13R-08  
 
==References:==
: 1. Letter from J. L. Hansen (Exelon Generation Company, LLC) to u. S. Nuclear Regulatory Commission, "Third 10-Year Inservice Inspection Interval Relief Request 13R-08, Alternative Requirements to ASME Section XI Appendix VIII (Supplements 2 and 10), Examinations of Class 1 Pressure Retaining Welds Conducted from the Inside Surface In Accordance with 10 CFR 50.55a(a)(3)(i)," dated April 11, 2011 2. Letter from J. L. Hansen (Exelon Generation Company, LLC) to u. S. Nuclear Regulatory Commission, "Supplemental Information Supporting Relief Request 13R-08: Alternative Requirements to ASME Code Requirements For Class 1 Pressure Retaining Welds," dated June 6, 2011 3. Letter from D. M. Gullott (Exelon Generation Company, LLC) to u. S. Nuclear Regulatory Commission, "Response to Request for Additional Information Related to Braidwood Station Relief Request 13R-08," dated November 2, 2011 4. Email from B. L. Mozafari (U. S. Nuclear Regulatory Commission) to L. A. Simpson (Exelon Generation Company, LLC), "Braidwood RAI regarding request for relief 13R-08, TAC Nos. ME6024 and ME6025," dated February 24,2012 (ADAMS Accession Number ML 12058A005)
In References 1, Exelon Generation Company, LLC, (EGC) submitted a request for relief, 13R-08, to the U. S. Nuclear Regulatory Commission (NRC) for review and approval.
This alternative requested relief from certain examination qualification requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, at Braidwood Station, Units 1 and 2. EGC proposed the use of root mean square (RMS) error criteria for sizing flaws that are greater than the requirements of ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds," and N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface." Reference 1 was supplemented by References 2 and 3.
March 12, 2012 U. S. Nuclear Regulatory Commission Page 2 In conference calls held January 12 and February 24,2012, the NRC indicated concerns with the submittals based on discussions with the industry.
By letter dated February 24,2012, the NRC requested additional information in support of the review (Reference 4). The Attachments provide the requested information.
The regulatory commitment contained in this letter is summarized in a table in Attachment
: 3. Should you have any questions concerning this letter, please contact Ms. Lisa A. Simpson at (630) 657-2815.
David M. Gullott Manager -Licensing Exelon Generation Company, LLC Attachments:
: 1) Response to Request for Additional Information
: 2) Supporting Information
: 3) Summary of Regulatory Commitments cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector
-Braidwood Station NRR Project Manager -Braidwood Station Illinois Emergency Management Agency -Division of Nuclear Safety ATTACHMENT 1 Response to Request for Additional Information NRC Position related to RMS Error Adjustment:
If twice the RMS error is not to be added to the depth of the measured flaw, the NRC staff requests that the licensee commit in their RAI response that flaw evaluations of any flaws found in the inspections covered by this relief request, 13R-OS, will be submitted to the NRC for review and approval prior to reactor startup. When submitting the evaluation, in addition to the typical information provided in a flaw evaluation, the following additional information will need to be included:
EGC Response to NRC Position related to RMS Error Adjustment:
Exelon Generation Company, LLC, (EGC) does not intend to use the NRC proposed RMS error adjustment.
EGC believes that the proposed adjustment is overly conservative.
The RMS error adjustment requested in Braidwood Relief Request (RR) 13R-OS, when combined with the margins required for evaluating flaws, provides reasonable assurance of the continued structural integrity of the subject welds. EGC has supported an industry assessment regarding this issue and has reviewed the statistical results. These results support the basis for the requested RMS error adjustment proposed in RR 13R-OS. Attachment 2 provides the industry assessment.
In lieu of using the staff proposed RMS error adjustment, EGC commits that flaw evaluations of detected flaws determined to be connected to the piping inner diameter surface during the examinations covered by Braidwood relief request 13R-OS will be submitted to the NRC, except as noted in the following responses (Reference Summary of Regulatory Commitments in Attachment 3 to this letter). NRC Additional Information Request No.1 related to Measured Flaw Size: The measured flaw size(s). EGC Response to NRC Request No.1 for Measured Flaw Size: If flaws are found, the measured flaw size as determined by UT will be included in the evaluation.
NRC Additional Information Request No.2 related to Proposed RMS Error Adjustment:
The RMS error that was added to the measured flaw size and how the RMS error was established.
The RMS depth-sizing error for the personnel conducting the examination.
The licensee may propose a smaller addition (less than twice the RMS error) to the flaw depth. A smaller addition to the flaw depth will need to be technically justified, which should include, as a minimum, a demonstration that the welds are easier to inspect than the Performance Demonstration Initiative Supplement 1010 specimens (e.g., no probe lift-off, less than 1/32-inch surface waviness) in the area near the flaw and the area used for depth-sizing.
Smooth 10 surfaces and other factors will be taken into account by NRC staff when reviewing the flaw evaluation.
Page 1 of 4 ATTACHMENT 1 Response to Request for Additional Information EGC Response to NRC Request No.2 related to Proposed RMS Error Adjustment:
The RMS error that was added to the measured flaw size and how the RMS error was established will be provided with the flaw evaluation.
The obtained procedure RMS sizing error provided in this relief request is a more appropriate value to use for an evaluation of a flaw through-wall measurement.
Providing the RMS value for a single individual does not provide additional assurance on the accuracy of the measurement.
Each analyst, while not qualified in accordance with Supplement 10, has demonstrated a capability of sizing to the same secondary acceptance criteria that is used to judge the procedure's performance.
Additionally, the data obtained to provide a through-wall depth measurement of reported flaws in the field is encoded and digitally stored. The data is essentially portable and can be sent easily to various analysts for concurrent flaw through-wall measurement evaluation.
Subsequently, the reported size of a flaw is not determined by a single analyst; the data is reviewed by several analysts and potentially outside consultants prior to supplying a final through-wall measurement determination used in the flaw analysis.
In addition, the individual depth sizing error performance is not provided to the utility or the vendor by the administrator of the qualification program. What is provided is the RMS error for the procedure and a list of candidates who have met or exceeded the procedure's RMS error performance value. It is EGC's understanding that the individual performance information could be made available to the NRC directly via the EPRI Performance Demonstration Initiative (POI) administrator, provided confidentiality controls are maintained; however, it is not in the best interest of the industry to have the individual results disclosed to a utility, its vendor, or the public, as it could compromise the integrity of the qualification program. In order to provide additional comparative information of the Braidwood Station's similar and dissimilar metal welds (OMWs) to be examined to the mockup welds, under which Supplements 2 and 10 qualification is conducted, the following was obtained from the POI administrator:
Procedures and personnel are qualified on realistic mock-ups that contain both field weld and shop weld configurations.
The shop weld configurations contain an 821182 dissimilar metal weld (Supplement
: 10) connecting a stainless steel-clad ferritic nozzle forging to a stainless steel safe-end, which is then joined using stainless steel metal material to either a wrought stainless steel pipe or a cast stainless steel elbow (Supplement 2 weld). The field weld configurations do not have safe-ends and connect the stainless steel-clad nozzle forging directly to either a wrought stainless steel pipe or a cast stainless steel elbow (Supplement 10 weld) using 82/182 weld material.
The Supplement 2 and Supplement 10 field welds contain the most aggressive geometry and are considered the most challenging; however, the location of these welds with relation to the relatively smooth shop weld also provide challenges to the techniques.
There are a total of six welds included in the POI qualifications.
Four of these welds are field welds (two dissimilar metal welds and two Supplement 2 welds) with similar geometric conditions and two dissimilar metal weld shop welds, one with a relatively long safe-end and the other with a short safe-end that brings the geometry from the field weld into the scan area. Page 2 of 4 ATTACHMENT 1 Response to Request for Additional Information Since the majority of the sample set is comprised of flaws that are contained in field weld configurations that have similar material and geometric conditions we feel the grading approach currently defined in the 2008 Addenda of Appendix VIII, Code Case (CC-696) and the POI qualification program continues to be technically appropriate.
Separating Supplement 2 measurements from the measurements made on similar weld configurations using the exact same techniques makes the evaluation more granular and reduces the overall confidence in the capabilities of the technique.
Given that one third of the combined Supplement 10 POI qualification welds are comparable to the Braidwood configuration it is reasonable to conclude that the OMWs for which relief is requested are less challenging to detect and size flaws versus the combination of geometries in the POI mockups. With regards to the Supplement 2 welds included in this relief, two-thirds of the similar and dissimilar metal welds included in the POI sample set are representative of the Braidwood welds so the combined RMS value of both Supplement 2 and 10 should be used. In summary, EGC believes the proposed adjustment to the depth size measurement can be technically justified for the range of conceivable flaws. NRC Additional Information Request No.3 related to Eddy Current Testinq: If the procedure uses eddy current, the determination by eddy current if the flaw is or is not surface breaking.
EGC Response to NRC Request No.3 related to Eddy Current Testinq: EGC will perform the required evaluations to determine if the flaw(s) are surface breaking or not. In the case of the examinations planned for spring 2012, the contracted examination vendor deploys eddy current in order to make these determinations.
NRC Additional Information Request No.4 related to Inner Diameter Profile of the Examination Area: The inner diameter profile of the weld, pipe, nozzle, and safe end (as applicable) in the region at and surrounding the transducer locations used to depth size the flaw. EGC Response to NRC Request No.4 related to Inner Diameter Profile of the Examination Area: If a flaw is detected and depth sizing is required, this information will be collected and provided with the evaluation.
NRC Additional Information Request No.5 related to Flaw Deqradation Mechanism:
The suspected flaw degradation mechanism and the process used to determine the degradation mechanism.
Page 3 of 4 ATTACHMENT 1 Response to Request for Additional Information EGC Response to NRC Reguest No.5 related to the Flaw Degradation Mechanism:
The initial nondestructive examination (NDE) data coupled with additional data collected will be used to aid in this determination and will be provided.
Page 4 of 4 ATTACHMENT 2 Supporting Information Industry Assessment Completed by Dominion Engineering, Inc. MRP 2012*011 r.:::::!'.!!r.:=t1211 EleCTRiC POWER .:,-RESEAI<CH INSTITUTE MRP Materials Reliability Program __________
M,RP 2012-011 (via email) March 8, 2012 To: MRP TAG, MRP IC, MRP Assessment TAC, and MRP Inspection TAC
 
==Subject:==
Inside Surface Flaw Depth Sizing Uncertainty Root Mean Square (RMS) Error Treatment Since 2002, the nuclear power industry has attempted to qualify personnel and procedures for depth-sizing examinations performed from the inside surface of dissimilar metal and austenitic stainless steel butt welds in PWR piping. To date, no domestic or international vendor has met the applicable root mean square (RMS) error requirement specified in the ASME Code. Utilities examining from the inner diameter have thus requested relief from the RMS error requirement by employing an adjustment of the measured flaw depth equal to the difference between the RMS error achieved in the qualification process and the RMS error required in ASME Section XI, Appendix VIII. This adjustment has been accepted by the NRC staff on numerous occasions, but recent staff review of qualification data has led to concerns with the adequacy of this adjustment.
A more conservative adjustment has been proposed by the NRC staff during the review of recent utility relief requests on this subject. The attachment to this letter presents an assessment of the procedure that has customarily been used by industry to account for the RMS error achieved during qualification for large-bore Alloy 82/182 and austenitic stainless steel butt welds in PWR piping. Also considered in this assessment is an alternative approach that was recently suggested by NRC staff. A simple statistical approach is taken to assess the effect of the alternative depth sizing approaches proposed by industry and NRC staff as compared to the ASME Section XI, Appendix VIII, RMS criteria.
Additionally, the implications of the depth sizing uncertainty on the use of stress improvement mitigation methods and on the disposition of flaws for continued service are specifically considered.
This letter report is being provided for use by member utilities in determining the most appropriate treatment of depth sizing uncertainty in upcoming inspection campaigns and is not considered proprietary information.
If you have any questions or concerns, please contact Craig Harrington (charrington@epri.com, 817-897-1433).
Together ... Shaping the Future of Electricity PALO ALTO OFFICE 3420 Hillview Avenue, Polo Alto, CA 94304-1338 USA. 650.855.2000
* Customer Service 800.313.377.4
* www.epri.com 20&#xa3;2 Best Regards, William Sims Entergy Chairman MRP Assessment TAC MRP 2012-011 Attachment 1: Assessment of Effect of the Depth Sizing Uncertainty for Ultrasonic Examinations from ID Surface of Large-Bore Alloy 82/182 and Austenitic Stainless Steel Butt Welds in PWR Primary System Piping Cc: Craig Harrington, EPRI Together ... Shaping the Future of Electricity PALO ALTO OFFICE 3420 Hillview Avenue, Polo Alto, CA 94304-1338 USA. 650.855.2000.
Customer Service 800.313.3774.
www.epri.com Dominion [nyineeriny, Inc ATTACHMENT 1 MRP 2012-011 Attachment 1, p. 1 of 20 ASSESSMENT OF EFFECT OF THE DEPTH SIZING UNCERTAINTY FOR ULTRASONIC EXAMINATIONS FROM ID SURFACE OF LARGE-BORE ALLOY 82/182 AND AUSTENITIC STAINLESS STEEL BUTT WELDS IN PWR PRIMARY SYSTEM PIPING 1 Introduction Since 2002, the nuclear power industry has attempted to qualify personnel and procedures for depth-sizing examinations performed from the inside surface of dissimilar metal and austenitic stainless steel butt welds in PWR piping. To date, no domestic or international vendor has met the applicable root mean square (RMS) error requirement of ASME Section XI Appendix VIII (Supplements 2 [4], 10 [5], and/or 14 [6]), or the alternative qualification requirements of ASME Code Case N-695 [7] or N-696 [8], as applicable.*
Utilities examining from the inner diameter have thus requested relief from the RMS error requirement by employing an adjustment of the measured flaw depth equal to the difference between the RMS error achieved in the qualification process and the RMS error required in Appendix VIII. This adjustment has been accepted by the NRC staff on numerous occasions, but recent staff review of qualification data has led to concerns with the adequacy of this adjustment.
A more conservative adjustment has been proposed by the NRC staff during the review of recent utility relief requests on this subject. The purpose of this document is to assess the procedure that has customarily been utilized by the industry to account for the RMS error achieved for large-bore Alloy 82/182 and austenitic stainless steel butt welds in PWR piping. The RMS error that is applied in this procedure is the RMS error achieved by the specific inspection vendor that performed the examination for the Appendix VIII supplement applicable to the examined weld. Also considered in this assessment is another alternative approach that was recently suggested by NRC staff. Specifically, the implications of the depth sizing uncertainty on the use of stress improvement mitigation methods and on the disposition of flaws for continued service are considered below A simple statistical approach is taken to assess the effect of the alternative depth sizing approaches proposed by industry and NRC staff .
* ASME Code Cases N-695 and N-696 have been approved by NRC for use without condition
[9].
Dominion [nyineeriny, Inc. 2 Background MRP 2012-011 Attachment 1 f p. 2 of 20 2.1 Requirements for Large-Bore Butt Welds in PWR Primary System Piping Not Fabricated with Alloy 82/182 ASME Section XI, Table IWB-2500-1, Examination Category B-F and Examination Category B-J, provides inspection requirements for visual, volumetric, and surface inspections of piping butt welds in the primary system that are not made of Alloys 82 and/or 182. Such welds are not considered to be susceptible to primary water stress corrosion cracking (PWSCC). Table IWB-2500-1 generally requires that the large-bore butt welds in the primary system piping not fabricated with Alloy 821182 be examined using volumetric and surface techniques during each IO-year in-service inspection interval.
However, subject to NRC approval, alternative inspection requirements may be applied for such locations on the basis of a risk-informed inspection program for piping implemented by the licensee (per ASME Section XI Appendix R). The volumetric examination requirements of Section XI including those for piping butt welds are addressed by ultrasonic examinations meeting the requirements of Section XI Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems." 2.2 Requirements for Large-Bore Butt Welds in PWR Primary System Piping Fabricated with Alloy 82/182 ASME Code Case N-770-1 [I] provides alternative inspection requirements for visual, volumetric, and surface inspections of piping butt welds in the primary system that are made of Alloys 82 and/or] 82, which are considered to be susceptible to PWSCC.
* The majority but not all of the dissimilar metal butt weld locations in PWR primary piping systems were fabricated using Alloy 82/182; stainless steel welds were used to join dissimilar base alloys in some cases. This code case has been made mandatory by the US NRC through regulation 10 CFR 50.55a(g)(6)(ii)(F), subject to the conditions detailed in this regulation.
The inspection requirements including inspection frequencies for Alloy 82/182 piping and nozzle butt welds were previously defined in Revision 1 ofMRP-139
[3]. MRP-139, Revision 1 and an ASME document [2] form the technical basis for the requirements ofN-770-1.
The volumetric examination requirements of N-770-1 are addressed by ultrasonic examinations meeting the requirements of Section XI Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems."
* An update ofN-770-1 (Code Case N-770-2, June 9, 2011) has been approved by ASME, but the version that is currently made mandatory by the NRC regulations will remain in effect until the next NRC final rule is issued in 2013 or 2014.
Dominion Inc MRP 2012-011 Attachment 1, p. 3 of 20 Code Case N-770-1 includes specific categories to address inspection methods and frequencies for piping Alloy 821182 dissimilar metal weld (DMW) locations both unmitigated and mitigated against PWSCC. N-770-1 includes Inspection Items A-I, A-2, and B for unmitigated welds and Inspection Items D and E to address butt welds mitigated with stress improvement with or without welding. Item D covers the case ofuncracked butt welds while Item E covers the case of cracked butt welds (i.e., with PWSCC type indications connected to the inside surface).
The two currently available stress improvement methods are the Mechanical Stress Improvement Process (MSIPTM), which is performed without welding, and Optimized Structural Weld Overlay (OWOL), which also credits reinforcement of the pressure boundary with PWSCC-resistant materiaL
* The NRC regulation 10 CFR 50.55a(g)(6)(ii)(F)(2) authorizes that "welds that have been mitigated by MSIPTM may be categorized as Inspection Items D or E, as appropriate, provided the [performance]
criteria in Appendix I of the code case have been met." Use of Inspection Items D or E for welds treated by OWOL currently requires application-specific review and approval by NRC. Code Case N-754 [25] was recently approved by ASME defining requirements for the design ofOWOLs, including the life of the overlay. For these stress improvement methods, the basic volumetric inspection requirement following mitigation of an uncracked DMW (Item D) is a single examination within 10 years following mitigation, t followed by a program of periodic inspections in which the component is placed into a population to be examined on a sample basis, provided that no indications of cracking are found. The basic volumetric inspection requirement following mitigation of a cracked DMW (Item E) is a single examination during the first or second refueling outage following application of stress improvement, followed by a program of periodic inspections in which the component is placed into a population to be examined on a sample basis, provided that no indications of crack growth or new cracking are found. :j:
* Water jet peening, fiber laser peening, and laser shock peening are additional mitigation methods under consideration that result in a layer of compressive residual stress at the wetted surface. t The NRC regulation 10 CFR SO.SSa(g)(6)(ii)(F)(9) modifies the timing of the follow-up examination to be "no sooner than the third refueling outage and no later than 10 years following stress improvement application." t The NRC regulation 10 CFR S0.5Sa(g)(6)(ii)(F)(8) adds the condition that ''welds mitigated by optimized weld overlays in Inspection Items D and E are not permitted to be placed into a population to be examined on a sample basis and must be examined once each inspection interval."
Dominion fnvineerinv, Inc. MRP 2012-011 Attachment 1, p. 4 of 20 2.3 Depth Sizing Error Requirement for Volumetric Examinations of Piping Butt Welds The depth sizing requirement for DMWs (including Alloy 82/182 butt welds) and austenitic stainless steel welds in PWR primary piping is defined in Appendix VIII of ASME Section XI using the RMS error for a performance demonstration:
[1 ] where RMS root mean square (RMS) error m i measured flaw size ti true flaw size n number of flaws measured The required RMS value is 0.125 inch per Appendix VIII (Supplements 2 [4], 10 [5], and 14 [6]), or the alternative requirements of ASME Code Case N-695 [7] or N-696 [8], as applicable.
Since 2002, the nuclear power industry has attempted to qualify personnel and procedures for this depth-sizing requirement for ultrasonic examinations performed from the inside surface of dissimilar metal and austenitic stainless steel butt welds in PWR piping. Four domestic and international inspection vendors have demonstrated a capability to depth-size flaws, but to date none of them has achieved an RMS error of 0.125 inch. Of the four vendors, the largest demonstrated flaw sizing RMS error for DMWs (i.e., Supplement
: 10) is 0.224 inch ([13], [33]).* Of the four vendors, the largest demonstrated flaw sizing RMS error for austenitic stainless steel piping welds (i.e., Supplement
: 2) is 0.367 inch [33]. It is noted that the required RMS error of 0.125 inch was originally based on the depth sizing error that was achievable for ultrasonic examinations of BWR piping welds in the 1980s, and that there is no specific technical requirement satisfied by the 0.125-inch error value .
* Of the four vendors, the largest demonstrated flaw sizing RMS error for Supplements 10 and 2 combined (Le., Supplement
: 14) is 0.245 inch [33]. This value is similar to, and less than 10% greater than, that for Supplement 10 alone (0.224 inch).
Dominion [nvineerinv, Inc. MRP 2012-011 Attachment 1, p. 5 of 20 2.4 Proposed Depth Sizing Procedures in Lieu of RMS Error Requirement of Appendix VIII Supplements 2, 10, and 14 and Code Cases N-695 and N-696 Given the impracticality of achieving the 0.125-inch RMS error value, the industry developed an alternative approach in which a quantity equal to the difference between the actual RMS error and an RMS error of 0.125 inch is added to the measured depth: m adj = m + (RMS -0.125 in.) [2] where m measured flaw size madj adjusted flaw size to be applied in flaw assessments RMS actual RMS error for applicable Appendix VIII supplement The intention of this proposed procedure is to bias the measured value upward to account for the increased measurement uncertainty versus an idealized examination satisfying the RMS error requirement.
This proposed alternative was submitted to the NRC by some individual licensees in relief requests (e.g., [10], [11], and [12] are for one such relief request).
In the past, utility relief requests proposing this alternative approach have been accepted by the NRC pursuant to 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(i) (e.g., [13]). However, recently the NRC staff has suggested another alternative approach to the depth-sizing issue in which a quantity equal to twice the actual RMS error is added to the measured depth [14]: m adj = m+2xRMS [3] Specifically, this alternative is suggested for qualification specimen diameters from 27 through 29 inches and wall thickness between 2.5 through 2.9 inches for Supplements 2 and 10. Using a simple statistical approach, the practical effects ofthese two proposed alternatives (equations
[2] and [3]) are assessed below. 3 Effect of Alternatives to RMS Depth Sizing Error Requirement of Section XI Appendix VIII Indications of flaws that are detected in DMWs and other butt welds in primary system piping must be dispositioned by repair, replacement, mitigation, or acceptance/evaluation for continued service. The replacement and often the repair option remove the indication from the subject component.
However, the mitigation and acceptance/evaluation options require the licensee to consider the depth of the flaw indication determined by NDE. The effect on the MSIPTM and Dominion Inc MRP 2012-011 Attachment 1, p. 6 of 20 OWOL mitigation methods is assessed in Sections 3.1 and 3.2, respectively.
The discussion in Sections 3.1 and 3.2 is specific to depth sizing of flaws connected to the inside surface of dissimilar metal butt welds (addressed by Supplement
: 10) because the MSlpTM and OWOL methods are used to mitigate PWSCC of Alloy 82/182 piping butt welds. The effect on acceptance/evaluation of indications of flaws detected in piping butt welds, especially unmitigated locations, for continued service is assessed in Section 3.3. In Section 3.3, the discussion is broadened to address the effect of depth sizing error in the context of dissimilar metal butt welds (addressed by Supplement
: 10) and wrought austenitic stainless steel butt welds (addressed by Supplement
: 2) in PWR piping. 3.1 Implications for Mechanical Stress Improvement Process (MSlp lM) to Mitigate Large-Bore Alloy 82/182 Dissimilar Metal Butt Welds in PWR Piping The MSIPTM method ([ 15] through [24]) was originally introduced in the nuclear power industry as a mitigation method for BWR piping subject to cracking mechanisms such as IOSCC. The MSIPTM method mitigates SCC by introducing a permanent compressive residual stress field on the inside surface of the DMW by way of mechanical squeezing.
The process redistributes the "as-welded" tensile residual stresses, resulting in compressive axial and hoop residual stresses on the ID surface extending to about the inner 50% of the wall thickness
([17], [18], and [20] through [24]). As a prerequisite for crediting MSIPTM mitigation, there is a standard requirement that an examination be performed showing that there are no crack indications on the ID surface deeper than 30% of the wall thickness or having a total circumferential extent greater than 10% of the circumference
([3], [15], [16], [17], and [19]). The requirement that any flaws have a depth no greater than 30% of the wall thickness ensures that such flaws are effectively mitigated by the process, considering factors such as the uncertainty in the flaw depth, the uncertainty in the depth of the compressive residual stress zone, and the effect of operating load stresses.
The practical effect of the two alternatives for adjustment of the measured depth is illustrated in Figure 1. This figure shows the cumulative distribution function (CDF),
* i.e., uncertainty distribution, for the true flaw depth under three different assumptions:
(1) a hypothetical UT depth sizing resulting in a measured flaw depth of30% of the wail thickness for an eXamination with an RMS error of 0.125 inch (labeled as "Code" in Figure 1),
* The cumulative distribution function of Figure 1 describes the probability that a flaw reported to have a depth from the ID of30% of the wall thickness (after any adjustment under the industry or NRC alternative) has in actuality a depth less than or equal to the value shown on the x-axis.
Dominion [nvineerinv, Inc MRP 2012-011 Attachment 1, p. 7 of 20 a) b) 1.0 0.9 = .S 0.8 ..... CJ = rZ 0.7 = 0.6 = ,.Q .t;: 0.5 ..... 11.1 .... aJ 0.4 ;;.-.s 0.3 = = 8 0.2 0.1 0.0 0% 1.00 = 0.95 Q .... ..... CJ = ::I 0.90 = Q .... ..... = ,.Q .t;: 0.85 ..... 11.1 .... Q aJ ;;.-0.80 eu -= = = U 0.75 (3) NRC Alternative.
Measurement uncertainty RMS error of 0.224 in. with twice actual RMS 2 xO .224 = 0.448 in.
___
.. ***,-********--*****;
(2) Industry Alternative.
Measurement .. __ uncertainty for RMS error of 0.224 in . with RMS error difference (0.224 -0.125) = 0.099 in. added to measured value ... " .. " .. " ... -." .... ...... _." .. Jl) Code. Measurement uncertainty
,_ ... ".,",." .. , for RMS error of 0.125 in. 10% 20% 30% 40% 50% 60% True Flaw Depth from ID (%wall) (3) NRC Alternative.
Measurement uncertainty for RMS error of 0.224 in. with twice actual RMS error 2xO.224 = 0.448 in. (2) Industry Alternative.
Measurement uncertainty for RMS error of 0.224 in. with RMS error difference (0.224 -0.125) = 0.099 in. added to measured value (1) Code. Measurement uncertainty for RMS error of 0.125 in. 0.70 -t--'--'--'---'--t--,---,---",---,--+JJ-->....JI"-'--f-'--'---'---'--t--'--'--'---'----;--.1..-J'---'.--'--l--'--'--'---"--l 20% 25% 30% 35% 40% 45% 50% 55% True Flaw Depth from ID (%wall) Figure 1. Supplement 10 Dissimilar Metal Butt Welds -Uncertainty for Flaw Depth Reported to Be 30%tw for 2.5-inch Wall Thickness: (a) Full Range of CDF, (b) Plot for CDF > 0.7 Dominion fnyineeriny, Inc MRP 2012-011 Attachment 1, p. 8 of 20 (2) a UT depth sizing for a process having the maximum RMS error of 0.224 inch, with a measured depth of 0.651 inch (26.04% of wall), resulting in an adjusted depth of 0.651 + 0.099 = 0.750 inch, or 30% of the wall thickness, per the industry-proposed alternative (Equation
[2]) (labeled as "Industry Alternative" in Figure 1), and (3) a UT depth sizing for a process having the maximum RMS error of 0.224 inch, with a measured depth of 0.302 inch (12.08% of wall), resulting in an adjusted depth of 0.302 + 2xO.224 = 0.750 inch, or 30% of the wall thickness, per the recent NRC alternative (Equation
[3]) (labeled as "NRC Alternative" in Figure 1). The upper portion (a) of Figure 1 illustrates the complete distribution function in each case, while the lower portion (b) of Figure 1 compares the upper tails of the distributions in greater detail. In Figure 1, the wall thickness is assumed to be 2.5 inches because this is the lower bound thickness in the range from 2.5 to 2.9 inches cited above [14], thus maximizing the relative depth error as a percentage of wall thickness.
In addition, the uncertainty in flaw depth is assumed to be normally distributed in each case, as is commonly assumed to describe measurement error. Finally, each distribution function shown in Figure 1 was truncated at a depth of 0% of the wall thickness, so the probability of the actual flaw depth being less than or equal to 0% is zero.
* This is a standard statistical approach that is applied when the assumed distribution extends beyond the range of physically meaningful values. The truncation was performed as follows: (fl d ) _ CDF( flaw depth)-CDF(O)
CDF;nmc aw epth CDF ( 0) [4] Thus, in each of the three cases, the reported flaw depth would be at the 30% limit of acceptability for mitigation by MSIPTM. A comparison of the second (2) or third (3) curve in Figure 1 with the first (1) curve illustrates how the adjustment in the reported depth tends to balance the effect of increased RMS error. For example, the adjustment of adding 0.099 inch to the measured depth effectively shifts the second (2) curve to the left by 4% ofthe assumed 2.5-inch wall thickness.
This shifting brings the upper tail of the second (2) curve into approximate alignment with the upper tail of the first (1) curve, with the two curves indicating the same cumulative probability level for an actual flaw depth of35% of the assumed wall thickness.
3.1.1 Assessment of Industry-Proposed Alternative A comparison ofthe curve for the industry-proposed alternative (2) with the hypothetical curve meeting the 0.125-inch RMS error requirement (1) shows that the industry-proposed alternative
* The truncation step in Figure 1 had a negligible effect on the first (l) and second (2) curves. The effect of the truncation on the third (3) curve was to reduce the cumulative probability at a depth of 0% from about 0.09 to zero. This had a negligible effect on the upper tail ofthe third (3) curve.
Dominion [nvineerinv, Inc. MRP 2012-011 Attachment 1, p. 9 of 20 is a reasonable approach in which most of the uncertainty distribution for the actual examination is conservatively bounded by the distribution for the idealized case meeting the RMS error requirement.
The upper tail ofthe distribution for the industry-proposed alternative extends only modestly beyond the upper tail for the idealized case, and there is an 84% probability that the industry-proposed approach produces a conservative result versus the idealized case meeting the Appendix VIII depth sizing RMS error requirement:
*,t where P[(tRMS -m adj) < (t o.125 -m) ] = p[[( m + Z(JRMS) -( m +(JRMS -(JO.l25)J
< [( m + Z(JO.l25)-
m JJ = p[ Z ( (J RMS -(JO.125) < ((J RMS -(J0.125) J =P[z<l] =0.84 m measured flaw size madj adjusted flaw size to be applied in flaw assessments tRMS true flaw size distribution per ultrasonic examination with actual RMS error [5] to. 125 true flaw size distribution per hypothetical ultrasonic examination with RMS error of 0.125 in. Z normal standard deviate aRMS true flaw size standard deviation per ultrasonic examination with actual RMS error aO.125 true flaw size standard deviation per hypothetical ultrasonic examination with RMS error of 0.125 in. = 0.125 in. In the context of an MSIPTM application to an Alloy 821182 DMW with crack indications, it is concluded that the alternative proposed by industry is an appropriate method to account for the impracticality of achieving the RMS error of 0.125 inch. Moreover, it is recognized that in the unlikely event that a flaw with an adjusted depth of 30% of wall were to have a true depth such that it was not effectively mitigated by MSIPTM, then it is highly likely that potential would be identified by the follow-up ultrasonic examination required by N-770-1 during the first or second refueling outage following the MSIPTM application.
That result would trigger flaw evaluation per IWB-3640 [30], as well as additional examinations during subsequent refueling outages or repair/replacement of the indication. Finally, it is also noted that, as shown in MRP-140 [32], leak-before-break behavior is predominant given circumferential cracking oflarge-bore PWR
* In this calculation, the same z-value is applied for the actual and idealized cases since the comparison is between an actual examination and its idealized case, and not between two distinct, independent examinations.
t As shown, the calculated probability of 0.84 is independent of the actual values for the RMS error for the actual and idealized examinations.
For an actual RMS error different than 0.224 inch, the calculated probability would still be 0.84 under the alternative proposed by industry.
Dominion fnvineerinv, Inc MRP 2012-011 Attachment 1, p. 10 of 20 piping because of its relatively high diameter-to-thickness ratio. Thus, in the unlikely event of extensive growth of the indication or indications sized prior to MSIPTM application, then there is high confidence the resulting leakage would be detected and acted upon while still maintaining a large margin against unstable flaw propagation.
3.1.2 Assessment of Recent Alternative Suggested by NRC Staff A comparison in Figure 1 of the curve for the recent alternative suggested by NRC staff (3) with the hypothetical curve meeting the O.l25-inch RMS error requirement (1) shows that the NRC alternative is clearly uncharacteristic ofthe distribution for the idealized case. For essentially all CDF values, the NRC alternative represents a large and overly conservative bias versus the idealized case meeting the RMS error requirement.
Similar to the above case for the proposed alternative, the probability that the recent NRC alternative produces a conservative result versus the idealized case is assessed as follows: p[ (tRMS -m adj ) < (t o.125 -m) ] = p[[( m + ZO"RMS )-( m + 20"RMS)J < [( m + ZO"O.125)-
m JJ = p[ Z (O"RMS -0"0.125) < 20"RMS J p[ 20"RMS] =
O"RMS -0"0.125 [6] For the case of the maximum actual RMS error for Supplement 10 of 0.224 inch, there is a 99.99970%
probability that the recently suggested NRC approach produces a conservative result versus the idealized case meeting the RMS error requirement:
p[ 20"RMS] Z < O"RMS -0"0.125 = p[Z < 2(0.224) ] 0.224-0.125
= p[Z < 0.448] 0.099 = p[ Z < 4.525] = 0.9999970
=1-3.0xlO-6
[7]
Dominion [nvineerinv, Inc MRP 2012-011 Attachment 1, p. 11 of 20 In the context of an MSIPTM application to a DMW with crack indications, it is concluded that the alternative recently suggested by NRC staff is unnecessarily conservative and inappropriate as a method to account for the impracticality of achieving the RMS error of 0.125 inch. The overly conservative nature of the NRC alternative would unnecessarily preclude the crediting of MSIPTM mitigation for indications that have a measured (pre-adjustment) depth as small as 12% ofthe wall thickness (for a wall thickness of2.5 inch). This 12% figure compares to 26% for the maximum allowable measured (pre-adjustment) depth for crediting ofMSIPTM mitigation under the industry alternative discussed above. This conclusion regarding the NRC alternative
[14] extends more generally to the situations for OWOL mitigation and disposition of flaws for continued service given their similarities to the situation for MSIPTM application as is apparent from the discussions below. 3.2 Implications for Optimized Structural Weld Overlay (OWOl) to Mitigate large-Bore Alloy 82/182 Dissimilar Metal Butt Welds in PWR Piping As introduced in Section 2 above, OWOL mitigation is another method that is available to mitigate PWSCC of Alloy 821182 DMW s in PWR primary system piping. Per NRC regulation, application-specific review and approval is required by NRC for the treated welds to be categorized as mitigated welds with regard to the inspection requirements ofN -770-1. The technical basis for OWOL mitigation, along with that for full structural weld overlay (FSWOL), is documented in MRP-169, Revision I-A [26], which was approved in 2010 by NRC [27] after an NRC-sponsored technical assessment including detailed modeling of the weld residual stresses associated with the OWOL process [28]. The OWOL mitigation credits the outer 25% of the wall thickness beneath the weld reinforcement in its structural design, and is effective through the combination of improved stress in the inner portion of the susceptible material and introduction ofPWSCC-resistant overlay material.
In 2011, Code Case N-754 [25] was approved by ASME defining requirements for the design of OWOLs, including the life ofthe overlay. N-754 includes requirements for the use ofOWOL as a "repair OWOL" in which the process is applied over material with flaws with a depth from the inside surface no greater than 50% of the pre-OWOL wall thickness.
* N-754 specifies that a crack growth calculation be performed to determine the life of the overlay based on the time for the detected flaws to grow to a depth of75% of the original pre-OWOL wall thickness.
This crack growth calculation considers the residual stresses that exist prior to application ofthe
* Under the industry alternative, the 50% through-wall flaw depth limit for repair aWOL corresponds to a measured (pre-adjustment) depth of about 46% for large-bore Alloy 82/182 piping butt weld locations.
Under the NRC alternative, the corresponding limit for the measured (pre-adjustment) depth is as small as 32% ofthe wall thickness.
Dominion fnvineerinv, Inc. MRP 2012-011 Attachment 1, p. 12 of 20 OWOL, and crack growth by both PWSCC and fatigue must be evaluated.
The analyzed life of the overlay is applied in Code Case N-770-2
* to limit the interval between volumetric examinations to the analyzed life, but no more than 10 years. Thus, the implications of uncertainty in the pre-OWOL flaw indication depth with regard to the effectiveness of OWOL mitigation are similar to the effect of uncertainty in the initial flaw depth for evaluations ofPWSCC flaws for continued service as discussed below in Section 3.3. As discussed below in Section 3.3, the standard ASME approach to crack growth calculations is to apply best-estimate type inputs except for the structural factors that are used to assess structural integrity for the end point of the crack growth calculation.
As shown above, the depth sizing adjustment proposed by the industry biases the best-estimate initial flaw depth so that there is an 84% probability that the industry-proposed approach produces a conservative result versus a hypothetical depth sizing meeting the requirement for an RMS error of 0.125 inch, and the uncertainty distribution for the true flaw depth per the industry-proposed alternative is reasonably characteristic of the uncertainty distribution for this hypothetical case. The flaw depth of75% of the pre-OWOL wall thickness defmes the end point of the crack growth calculation of overlay life, meaning that at the end of overlay life the predicted flaw depth remains outside of the outer 25% of the original wall credited in the OWOL structural design. Furthermore, there is a requirement in N-754 that the OWOL design exhibit minimum structural factors albeit reduced from the full standard ASME structural margins under the assumption of circumferential cracking extending around the entire circumference of the item and 100% through the susceptible material.
Given these conservatisms inherent in the OWOL design and the standard ASME approach of using the best-estimate initial flaw size as input to crack growth calculations, it is concluded that the alternative approach proposed by the industry is appropriate to address the impracticality of meeting the required depth sizing RMS error of 0.125 inch. Moreover, it is recognized that both N-770-1 and N-770-2 require a follow-up ultrasonic examination of the treated item during the first or second refueling outage following the OWOL application.
If crack growth is detected during this follow-up examination, then additional actions are required such as applying flaw acceptance standards and performing repeat volumetric examinations during multiple refueling outages. Thus, this follow-up examination requirement is another significant source of conservatism with regard to repair OWOL.
* AS ME Code Case N-770-2 was approved by ASME on June 9,2011, but N-770-1 is the version currently made mandatory by the NRC regulations.
Dominion fnyineeriny, Inc MRP 2012-011 Attachment 1, p. 13 of 20 3.3 Implications for Disposition of Flaws Detected in Large-Bore Dissimilar Metal and Wrought Austenitic Stainless Steel Butt Welds in PWR Piping The majority but not all ofthe dissimilar metal butt weld locations in PWR primary piping systems were fabricated using Alloy 82/182, with stainless steel welds used to join dissimilar base alloys in some cases. Unlike stainless steel weld material, Alloy 82/182 welds are susceptible to PWSCC. Thus, the flaw disposition procedures of Section XI require that planar surface-connected flaws that are in contact with the reactor coolant and are detected in Alloy 82/182 weld material be evaluated considering growth due to fatigue and PWSCC. For such flaws detected in stainless steel weld material, growth due to fatigue only must be considered.
Hence, flaw disposition is assessed separately below for dissimilar metal (Supplement
: 10) piping welds, including Alloy 82/182 welds, and for wrought austenitic (Supplement
: 2) piping welds, which were fabricated using stainless steel weld materiaL 3.3.1 Disposition of Flaws Detected in Large-Bore Dissimilar Metal Butt Welds in PWR Piping (Supplement
: 10) The required procedure for evaluation and acceptance of planar surface-connected flaws in contact with the reactor coolant environment in large-bore Alloy 82/182 dissimilar metal butt welds is defined by ASME IWB-3640 [30]. In this procedure the flaw size at the end of the assumed evaluation period is calculated based on deterministic equations of SCC and fatigue crack growth. The acceptable flaw size at the end of the assumed evaluation period is determined through a flaw stability calculation in which structural factors greater than one are applied to operating loads, and the end-of-evaluation-period flaw depth is limited to 75% ofthe wall thickness.
In this deterministic approach, best-estimate type inputs including for the initial flaw size based on NDE are used except for the use of structural factors on the operating loads. The conservative nature of the flaw disposition procedure is due to the use ofthe structural factors and the 75% limiting flaw depth. In the particular case of the PWSCC crack growth rate equation recommended in C-8511 of Section XI for evaluation of flaws in Alloy 821182 butt welds, this deterministic crack growth rate equation was developed in MRP-115 [31] to bound the log-mean behavior of 75% of the test welds included in the worldwide set oflaboratory data considered.
The 75 th percentile was chosen in MRP-115 in recognition that welds showing a higher crack growth rate than average (normalized for temperature, loading, and environment) are also more likely to initiate flaws. As shown above, the depth sizing adjustment proposed by the industry biases the best-estimate initial flaw depth so that there is an 84% probability that the industry-proposed approach produces a conservative result versus a hypothetical depth sizing meeting the requirement for an Dominion Inc MRP 2012-011 Attachment 1, p. 14 of 20 RMS error of 0.125 inch. Given that best-estimate type inputs except for the structural factors are used in the ASME procedure, it is concluded that the industry-proposed alternative approach is appropriate for dissimilar metal piping welds including those fabricated using Alloy 82/182 to address the impracticality of meeting the required depth sizing RMS error of 0.125 inch. Moreover, it is recognized that in the unlikely event that the actual end-of-evaluation-period flaw size were to exceed the size calculated in the flaw evaluation, then the result with high probability would be a stable flaw deeper than 75% of the wall thickness or a stable through-wall flaw detected via evidence ofleakage.
This conclusion is supported by MRP-140 [32], which demonstrates that leak -before-break behavior is predominant given circumferential cracking of large-bore PWR piping because of its relatively high diameter-to-thickness ratio.
* 3.3.2 Disposition of Flaws Detected in Large-Bore Wrought Austenitic Stainless Steel Butt WeLds in PWR Piping (SuppLement
: 2) The above discussion also generally applies to the case of disposition of flaws detected in bore austenitic stainless steel butt welds in PWR piping. In this case, PWSCC crack growth does not apply and ASME IWB-3514 [29] can be used to accept relatively shallow planar flaws that are in contact with the reactor coolant. However, ID surface-connected planar flaws that are deeper than permitted by IWB-3514.1 that are left in service must be evaluated using IWB-3640.
Again, the conservatism in the procedure is due to the use ofthe structural factors and the 75% limiting flaw depth. Other inputs to the procedure including the initial flaw size based on NDE are generally best-estimate type inputs. Similar to Figure 1, Figure 2 shows the depth sizing uncertainty distributions based on the maximum demonstrated flaw sizing error for Supplement 2 (0.367 inch [33]) for the same three cases considered in Figure 1. In Figure 2, the point of intersection between the curves representing the industry alternative (2) and the idealized case (1) is at a cumulative probability level of about 84%. This is the same cumulative probability value as for the intersection point in Figure 1 for these two cases because this point of intersection is independent of the actual RMS error value as shown in equation [5].t
* MRP-140 presents calculation results for PWSCC crack growth of through -wall circumferential flaws. The limiting case within the set of large-bore Alloy 82/182 piping butt welds is for a reactor vessel outlet nozzle. For this case, a period of 11.9 years is calculated for growth from a circumferential length corresponding to the technical specification leak rate limit of 1 gpm to the critical flaw length. t As for Figure 1, the distributions in Figure 2 were truncated for depths below 0% of wall thickness.
This resulted in a modest shifting of the second (2) curve in Figure 2, reducing the cumulative probability at a depth of 0% from about 0.08 to zero. This shifting of the second (2) curve lowered the actual intersection point with the first (1) curve slightly down to a probability of82%. The truncation for the third (3) curve in Figure 2 reduced the cumulative probability at a depth of 0% from about 0.48 to zero.
Dominion Inc MRP 2012-011 Attachment 1, p. 15 of 20 a) b) 1.0 (3) NRC Alternative.
Measurement uncertainty 0.9 for RMS error of 0.367 in. with twice actual RMS error 2 x 0.367 = 0.734 in. __ ____ ,_,_,_, __ , ______ , ________ , __ added to measured value 0% 1.00 10% 20% 30% (2) Industry Alternative.
Measurement
----'uncertainty for RMS error of 0.367 in. with RMS error difference (0.367 -0.125) = 0.242 in. added to measured value Code. Measurement uncertainty for RMS error of 0.125 in. 40% 50% 60% True Flaw Depth from ID (%wall) (3) NRC Alternative.
Measurement uncertainty for RMS error of 0.367 in. with twice actual RMS error 2xO.367 = 0.734 in. added to measured value (2) Industry Alternative.
Measurement uncertainty for RMS error of 0.367 in. ,_',_w __ ", ______ w,w,,, with RMS error difference (0.367 -0.125) = 0.242 in. added to measured value (1) Code. Measurement uncertainty for RMS error of 0.125 in. 0.70 +-"--'--k--'--!--'---'---'----"'--t--'-...!..4-'--'-t--'--'--'--'--t--'--'--'--'-+-"--'---"---'--!---'---'--'---"--I 20% 25% 30% 35% 40% 45% 50% 55% True Flaw Depth from ID (%wall) Figure 2. Supplement 2 Wrought Austenitic Stainless Steel Butt Welds -Uncertainty for Flaw Depth Reported to Be 30%tw for 2.5-inch Wall Thickness: (a) Full Range of CDF, (b) Plot for CDF > 0.7 Dominion tnvineerinVr Inc MRP 2012-011 Attachment 1, p. 16 of 20 Thus, as in the case of Supplement 10, the industry alternative in the case of Supplement 2 results in an uncertainty distribution for flaw depth that with a probability of about 84% bounds the idealized case meeting the Appendix VIII depth sizing error requirement.
Comparing Figure 1 and Figure 2, the upper tail for the largest achieved RMS error for Supplement 2 extends further toward greater depths than that for Supplement
: 10. This is judged to be acceptable recognizing that the Supplement 2 welds are not susceptible to PWSCC, and that flaw growth by fatigue is generally small in comparison to that by PWSCC in the Alloy 600/82/182 materials that are susceptible to PWSCC (see, e.g., [34] and [35]). Hence, it is concluded that the alternative approach proposed by the industry is also appropriate for wrought austenitic Supplement 2 piping welds to address the impracticality of meeting the required depth sizing RMS error of 0.125 inch. Figure 2 also facilitates a comparison of the effect of the NRC alternative (3) versus the idealized case (1) given the largest demonstrated RMS error for Supplement
: 2. For nearly all CDF values, the NRC alternative represents a large and overly conservative bias versus the idealized case meeting the RMS error requirement.
As calculated using equation [6], for the case of the maximum actual RMS error for Supplement 2 of 0.367 inch, there is a 99.88% probability that the recently suggested NRC approach produces a conservative result versus the idealized case meeting the RMS error requirement:
* pi z < 2 aRMS ] L (JRMS -a O.l 25 = p[z < 2(0.367) ] 0.367 -0.125 = p[z < 0.734] 0.242 = p[ z < 3.033] = 0.9988 [8] In the context of disposition of flaws in both Supplement 2 and 10 piping welds, it is concluded that the alternative recently suggested by NRC staff is unnecessarily conservative and inappropriate as a method to account for the impracticality of achieving the RMS error of 0.125 inch. It is noted that the under the NRC alternative, the adjustment to the measured flaw depth may be as large as about 29% of the wall thickness, compared to as large as about 10% of the wall thickness under the industry alternative .
* The truncation step in Figure 2 caused a slight shifting of the upper tail of the third (3) curve, lowering the actual intersection point ofthe third (3) and first (1) curves in Figure 2 to a probability of about 99.7% rather than 99.88%.
Dominion fnvineerinVr Inc 4 Conclusion MRP 2012-011 Attachment 1, p. 17 of 20 Compliance with the 0.125-inch depth sizing RMS error required by ASME Code Section XI Appendix VIII (Supplements 2, 10, and 14), or the alternative requirements of ASME Code Case N-695 or N-696, as applicable, is impractical for ultrasonic examinations from the ID surface. The alternative proposed by the industry to add the difference between the required RMS error value of 0.125 inch and the actual RMS error value for the selected inspection vendor, up to the maximum demonstrated RMS value of 0.224 inch, in conjunction with the use of appropriate acceptance standards, continues to provide reasonable assurance of structural integrity of the subject welds. In summary, the alternative which has been customarily used is an appropriate means of addressing the impracticality of the RMS error requirement for large-bore Alloy 82/182 and austenitic stainless steel butt welds in PWR piping. The alternative recently suggested by NRC staff of adding twice the applicable RMS error to the measured depth is unnecessarily conservative as clearly seen by comparison ofthe depth size uncertainty distribution for this alternative with that for the idealized case meeting the Appendix VIII depth sizing error requirement.
While the NRC approach would grossly mischaracterize flaw depths in an effort to address the actual RMS error achieved, the industry proposal conservatively treats the large majority of indications without unnecessarily distorting the measured flaw depth. 5 References
: 1. ASME Code Case N-770-1, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities," Section XI, Division 1, American Society of Mechanical Engineers, New York, Approved December 25,2009. 2. P. Donavin, G. Elder, and W. Bamford, "Technical Basis Document for Alloy 82/182 Weld Inspection Code Case N-770 and N-770-l," dated July 27, 2009. 3. Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139, Revision 1), EPRI, Palo Alto, CA: 2008. 1015009. 4. 2007 ASME Boiler & Pressure Vessel Code, Section XI, Mandatory Appendix VIII Performance Demonstration for Ultrasonic Examination Systems, Supplement 2 Qualification Requirements for Wrought Austenitic Piping Welds, 2008a Addenda, American Society of Mechanical Engineers, New York, July 1, 2008. 5. 2007 ASME Boiler & Pressure Vessel Code, Section XI, Mandatory Appendix VIII Performance Demonstration for Ultrasonic Examination Systems, Supplement 10 Qualification Requirements for Dissimilar Metal Piping Welds, 2008a Addenda, American Society of Mechanical Engineers, New York, July 1, 2008.
Dominion [nyineeriny, Inc MRP 2012-011 Attachment 1, p. 18 of 20 6. 2007 ASME Boiler & Pressure Vessel Code, Section XI, Mandatory Appendix VnI Performance Demonstration for Ultrasonic Examination Systems, Supplement 14 Qualification Requirements for Coordinated Implementation of Supplements 10, 2, and 3 for Piping Examinations Performed from the Inside Surface, 2008a Addenda, American Society of Mechanical Engineers, New York, July 1, 2008. 7. ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds," Section XI, Division 1, American Society of Mechanical Engineers, New York, Approved May 21, 2003. 8. ASME Code Case N-696, "Qualification Requirements for Appendix VIn Piping Examinations Conducted from the Inside Surface," Section XI, Division 1, American Society of Mechanical Engineers, New York, Approved May 21,2003. 9. USNRC, Regulatory Guide 1.147, Revision 16, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," October 2010. 10. Letter from M. J. Ajluni (Southern Nuclear Operating Company, Inc.) to US NRC, "Joseph M. Farley Nuclear Plant Unit 1 Proposed Relief Request for the Fourth lSI Interval ISI-RR-Ol)," NL-ll-0463, dated March 28,2011, ADAMS Accession No. MLll0871951.
: 11. Letter from R. Martin (USNRC) to M. 1. Ajluni (Southern Nuclear Operating Company, Inc.), "Joseph M. Farley Nuclear Plant, Unit 1 -Request for Additional Information on Inservice Inspection Plan (TAC NO. ME5966)," dated July 14,2011, ADAMS Accession No. MLll173A047.
: 12. Letter from M. J. Ajluni (Southern Nuclear Operating Company, Inc.) to US NRC, "Joseph M. Farley Nuclear Plant -Unit 1 Response to NRC Request for Additional Information
-Proposed Relief Request FNP-ISI-RR-Ol," NL-11-1543, dated August 11, 2011, ADAMS Accession No. ML112232241.
: 13. Letter from H. Chemoff(USNRC) to P. Freeman (NextEra Energy Seabrook, LLC), "Seabrook Station, Unit No.1 -Relief Request for Use of Alternate Depth Sizing Qualification, Third lO-Year Interval (TAC NO. ME3623)," with Enclosure "Safety Evaluation by the Office of Nuclear Reactor Regulation Relief Request Associated with Depth Sizing Acceptance Criteria NextEra Energy Seabrook, LLC Seabrook Station, Unit No.1 Docket No. 50-443," dated November 22,2010, ADAMS Accession No. ML103190139.
: 14. USNRC, "NRC Recommended Interim Approach to Address ASME Section XI, Appendix VIn Supplements 2 and 10 Inside Diameter Depth Sizing Root Mean Square Error (RMSE)." 15. USNRC, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping (Generic Letter 88-01)," dated January 25, 1988. 16. USNRC, Technical Report on Material Selection and Processing Guidelines for B WR Coolant Pressure Boundary Piping, NUREG-0313, Revision 2, January 1988, ADAMS Accession No. ML031470422.
Dominion Inc MRP 2012-011 Attachment 1, p. 19 of 20 17. Materials Reliability Program: Mechanical Stress Improvement Process (MSIP) Implementation and Performance Experiencefor PWR Applications (MRP-121), EPRI, Palo Alto, CA: 2004.1009503.
: 18. G. Rao, E. Ray, M. Badlani, and T. Damico, "An Assessment ofMSIP for the Mitigation ofPWSCC in PWR Alloy 82/182 Thick Wall Piping Welds by Instrumented Full Scale Mockup Testing," Proceedings of the Vessel Penetration Conference on Inspection, Crack Growth and Repair, US NRC, NUREGICP-0191 , September 2005. 19. USNRC, Memo from B. Sheron to E. Leeds, "Evaluation of the Mechanical Stress Improvement Process as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized Water Reactors," dated November 02,2009, ADAMS Accession No. ML092990640.
: 20. L. Fredette and P. Scott, Battelle Columbus, Evaluation of the Mechanical Stress Improvement Process (MSIP) as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized Water Reactors, September 2009, ADAMS Accession No. ML092990646.
: 21. L. Fredette, Pacific Northwest National Laboratory, Surge Nozzle NDE Specimen Mechanical Stress Improvement Analysis, NRC JCN N6319, PNNL-20549, July 2011, ADAMS Accession No. MLll1960318.
: 22. Westinghouse, Salem Unit 1 Reactor Vessel Outlet Nozzle Dissimilar Metal Weld Flaw Evaluation (Fall 2008 Outage), LTR-PAFM-08-158, December 2008, Westinghouse Proprietary Class 3, ADAMS Accession No. ML090500369.
: 23. M. Badlani, D. Bhowmick, and W. Bamford, "MSIP, A Proven Technique to Enhance Leak Before Break in Susceptible Reactor Vessel Nozzle Welds," ADAMS Accession No. ML080240480.
: 24. G. Elder, "MSIP Overview," ADAMS Accession No. ML082190517.
: 25. ASME Code Case N-754, "Optimized Structural Dissimilar Metal Weld Overlay for Mitigation of PWR Class 1 Items," Section XI, Division 1, American Society of Mechanical Engineers, New York, Approved June 25,2011. 26. Materials Reliability Program: Technical Basis for Preemptive Weld Overlays for Alloy 821182 Butt Welds in Pressurized Water Reactors (PWRs) (MRP-169)
Revision I-A, EPRI, Palo Alto, CA: 2010. 1021014. 27. US NRC, "Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report "Materials Reliability Program (MRP): Technical Basis for Preemptive Weld Overlays for Alloy 82/182 Butt Welds In PWRs (MRP-169)" Project Number 689," ADAMS Accession No. ML101660468.
: 28. L. Fredette and P. Scott, Battelle Columbus, Evaluation of Full Structural and Optimized Weld Overlays As Mitigation Strategies for Primary Water Stress Corrosion Cracking in Pressurized Water Reactors, April 2010, ADAMS Accession No. ML101260540.
: 29. 2007 ASME Boiler & Pressure Vessel Code, Section XI, IWB-3514 Standards for Examination Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles, Dominion fnyineeriny, Inc MRP 2012-011 Attachment 1, p. 20 of 20 and Examination Category B-J, Pressure Retaining Welds in Piping, 2008a Addenda, American Society of Mechanical Engineers, New York, July 1,2008. 30. 2007 ASME Boiler & Pressure Vessel Code, Section XI, IWB-3640 Evaluation Procedures and Acceptance Criteria for Flaws in Austenitic and Ferritic Piping, 2008a Addenda, American Society of Mechanical Engineers, New York, July 1, 2008. 31. Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696. 32. Materials Reliability Program: Leak-Before-Break Evaluation for PWR Alloy 821182 Welds (MRP-140), EPRI, Palo Alto, CA: 2005. 1011808. 33. Email from C. Latiolais (EPRI) to G. White (DEI), "RMS Sizing Error for ID Nozzle Exams," dated February 24,2012. 34. Materials Reliability Program: Alloy 821182 Pipe Butt Weld Safety Assessment for u.s. PWR Plant Designs (MRP-I09):
Westinghouse and CE Design Plants, EPRI, Palo Alto, CA: 2005. 1009804. 35. Materials Reliability Program: Alloy 821182 Pipe Butt Weld Safety Assessmentfor US PWR Plant Designs: Babcock & Wilcox Design Plants (MRP-112), EPRI, Palo Alto, CA: 2004. 1009805.
ATIACHMENT3 Summary of Regulatory Commitments The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)
COMMITMENT TYPE COMMITMENT COMMITTED DATE OR ONE-TIME PROGRAMMATIC "OUTAGE" ACTION ACTION (Yes/No) (Yes/No) EGC commits that flaw evaluations of detected Prior to A 1 R 16 flaws determined to be connected to the piping (spring 2012) reactor inner diameter surface during the examinations startup for covered by Braidwood relief request 13R-08 will Braidwood Unit 1. be submitted to the NRC, except as described in Prior to A2R 16 EGC letter RS-12-040 dated 3/12/2012. (fall 2012) reactor Yes No Applicable to Braidwood Station, Units 1 and 2. startup for Braidwood Unit 2.}}

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