ML071230013
| ML071230013 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/02/2007 |
| From: | American Electric Power Co |
| To: | Office of Nuclear Reactor Regulation |
| Vaidya B, NRR/DORL/LP4, 415-3308 | |
| Shared Package | |
| ML071230009 | List: |
| References | |
| N-716, TAC MD3137, TAC MD3138 | |
| Download: ML071230013 (21) | |
Text
Responses to NRC RAI on D C Cook Nuclear Plant (CNP)
Risk Informed ISI Application per ASME Section XI Code Case N-716
PART 1 - First Set of Questions (1)
The licensee requests authorization to implement a risk-informed inservice inspection (ISI) program based on American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code Case N-716 (N-716)." There appears to be, however, some differences between the methodology in N-716 and the method applied by the licensee:
(a)
Table 3 of N-716 discusses high, medium, and low failure potential, and pairs these potentials with degradation categories large break, small leak, and none respectively. It does not appear that this table was used in the submittal. Was this table used in the submittal? If not, what was used in lieu of Table 3?
Response
The information contained in Table 3 of N-716 was used in the CNP application and submittal. The information is identified in Tables 3.4-1/3.4-2 and Tables 5-1/5-2 of the submittal. The information is contained in the column identified as Failure Potential.
This column is further divided into two sub-columns (i.e. DMs and Rank). The Failure Potential Rank for HSS locations is then assigned as high, medium or low depending upon potential susceptibly to the various types of degradation. [Note: LSS locations were conservatively assumed to be a rank of medium (i.e. Assume Medium). See response to question 4b.
(b)
Section 5(c) of N-716 does not appear to provide a "with probability of detection (POD)" and "without POD" option in the calculation, but the submittal includes one set of estimates for "with POD" and another "w/o POD" in Table 3.4-1. Please clarify how the "with POD" and "w/o POD" columns in Table 3.4-1 are consistent section 5(c) of N-716.
Response
It is true that N-716 does not discuss the two options presented above. The CNP submittal contained both options in order to be consistent with previous RI-ISI submittals which contained both options. These two sets of analyses are typically conducted to provide a sensitivity of the delta risk evaluation with respect to assumptions on POD.
(c)
The estimates in the "w/o POD" column in Table 3.4-1 all seem to include a standard POD of 0.5. Is this correct? If not, please provide some examples using the conditional core damage probability (CCDP) values from page 11 of 35 to produce the entries in Table 3.4-1 and 3.4-2.
Response
That is correct. The w/o POD column applies a POD of 0.5 for both the Section XI program and the N-716 program. Thus, there is no extra credit assumed for an N-716 inspection as compared to Section XI inspection as to inspection effectiveness (e.g. due to larger inspection volumes in the N-716 program).
(d)
Section 7 of N-716, "Program Updates," includes several steps that make up a program update. Page 15 of 35 in the licensee's submittal states that, "[u]pon approval of the RIS-B Program, procedures that comply with the guidelines described in Reference 2 [ Electric Power Research Institute (EPRI) TR-112657 (EPRI Topical)] will be prepared to implement and monitor the program." Please identify the Sections in the EPRI topical that describe the update program that the licensee intends to implement. Please describe and compare the update program that the licensee intends to implement against the characteristics of such a program as described in Section 7 of N-716.
Response
The wording in the CNP submittal was based on previous RI-ISI submittals. While the intent of both updating processes (EPRI TR-112657 and N-716) is the same, I&M will meet the wording of N-716.
(2)
Regulatory Guide (RG) 1.178, "An Approach for Plant-Specific Risk-Informed Decision making for Inservice Inspection of Piping," describes one acceptable process for developing a RI-ISI program. Please explain how:
(a)
The approach used to analyze piping system failures for the plant-specific PRA of pressure boundary failures compares to the approach described in Section 2.1.4 of RG 1.178;
Response
The purpose of segments and segment definitions are identical between the ASME Code Case N-716 (N-716) approach and that of the EPRI RI-ISI methodology. In both methodologies, segments are used only as an accounting/tracking tool. That is, whether the weld is tracked individually or as part of a segment, the results of the risk ranking and element selection part of the methodology will not change. In both approaches, whether the segment is small (e.g., a single weld) or large (e.g., many welds), all of the welds will be ranked and then subject to a fixed sampling percentage for determining the size of the inspection population.
As an example, if the population of high safety significant (HSS) welds is 100, whether they are tracked as ten (10) segments (e.g., ten welds per segment) or two (2) segments (50 welds per segment), all 100 welds would be subject to the element selection process.
For example, 25% of HSS welds with susceptibly to a degradation mechanism would be selected for N-716 applications and 25% of welds identified as Risk Category 2 would be selected for EPRI RI-ISI applications.
(b)
The process used to assess piping failure potential for the plant-specific PRA of pressure boundary failures compares to the process outlines in section 2.1.5 of RG 1.178;
Response
For N-716 applications, failure potential is used in two ways:
- Confirm on a plant-specific basis that there is no other piping that should be considered as HSS per Section 2(a) of N-716. [Please see the response to Question 2(c) and 6(c) below.]
- Once the HSS population has been determined for the plant, the failure potential evaluation is identical to that in EPRI TR-112657 as applied to a number of NRC-approved RI-ISI applications. That is, the degradation mechanisms assessed, the evaluation criteria (e.g. attributes such as operating temperatures, allowable delta Ts, susceptible materials, flow velocities, etc.) and the failure potential ranking are the same.
(c)
The quantitative results of the pipe failure frequency that resulted from the failure potential assessment compares to the weld failure frequencies proposed in Section 5(a) of N-716 that are eventually used in your change in risk estimates;
Response
Because the failure frequencies in Section 5(a) of N-716 are at the weld level, they are, therefore, substantially smaller than what is used in conducting an internal flooding study in general, and the CNP internal flooding study, in particular. Another reason the failure frequencies used in the CNP internal flooding study are larger than the values used in the N-716 application is because the CNP internal flooding study includes the impact of flood sources beyond piping (e.g., tanks, pumps, heat exchangers, etc.). For screening purposes, this is conservative from an internal flooding study perspective.
(d)
The consequence evaluation performed as part of the plant-specific PRA of pressure boundary failures compares with the process outlined under Section 2.1.6 of RG 1.178.
Deleted: of Deleted:
Deleted: It is also conservative from a N-716 perspective because some of these flooding sources and therefore their contribution to failure frequency (e.g.,
tanks) are not within the N-716 scope of application (i.e., piping).
Response
The plant-specific PRA of pressure boundary failures is consistent with that discussed in Section 2.1.6 of RG 1.178 in that plant walkdowns were conducted to identify flood initiators and the locations of critical components. Additionally, for each flood zone and/or scenario, the impact of both direct and indirect effects was considered. Direct effects included loss of a train or system (e.g., loss or diversion of flow), an initiating event, or both. Indirect effects included spatial effects such a spray, pipe whip, etc. as well as loss of inventory effects (e.g., loss of a common tank).
(3)
Please fully define the population of welds to which the 10% guideline is applied and what inspections are counted.
(a)
Is the guideline to examine a minimum 10% of all high-safety-significant (HSS) welds, 10% of all HSS butt welds, 10% of all HSS butt welds >= 4 NPS, or something else?
Response
Yes, the guideline is to examine a minimum of 10% of HSS welds. For CNP, this population includes welds that are both less than, equal to, and greater than 4 NPS. It also includes butt welds and sockets welds.
Additionally, a lessons learned from the CNP application was that the wording of N-716 could be clearer in its intent to require inspection of at least 10% of the reactor coolant pressure boundary (RCPB). While the CNP application meets this intent, it is also the authors intent to revise N-716 to make this requirement clearer, as well as other lessons learned from N-716 applications [Please see response to Question 4(a) below.
(b)
What type of inspections can be counted, e.g., can visual examinations or wall thickness exams be counted in the 10%?
Response
Per N-716, wall thickness exams as part of the FAC and localized corrosion (excluding crevice corrosion) programs cannot be counted as part of the 10% required population.
Because of the nature of the degradation, wall thinning examination for locations potentially susceptible to erosion-cavitation will be conducted.
Per N-716, the requirements for examination of socket welds and smaller bore branch connections (i.e., < 2 NPS) susceptible to thermal fatigue shall be a volumetric exam of the piping base metal within 1/2 inch of the toe of the weld and a visual of the fitting itself. This is consistent with the requirements of EPRI MRP-146. The MRP-146 Deleted: CNP PRA to verify. Yes, t Inserted: Yes, t Deleted: see the response to Deleted: from the first set of RAIs]
evaluation has shown no small bore piping susceptible to swirl penetration thermal fatigue at CNP.
Thus, HSS inspections required by N-716 shall be volumetric exams as part of the CNP application (c)
What percentage of Class 1 butt welds (regardless of normal pipe size (NPS)) will be inspected in the proposed risk-informed program?
Response
I&M has selected a 14.7% sample for Unit 1 and a 10.1% sample for Unit 2 of Class 1 butt welds for examination regardless of NPS.
(4)
Section 5(c) in N-716 does not clearly specify what population of welds should be included in the change of risk estimates and what welds may be excluded. The description of the parameters in the equations in Section 5(c) indicates that any weld that was inspected under Section XI or that will be inspected under the RI-ISI program will be included in the Change in risk estimate.
(a)
Is the population of welds that should be included in theN-716 change in risk estimate all welds that were inspected under Section XI, and that will be inspected under the RI-ISI program? If not, where in code Case N-716 is the guidance that reduces the population of welds that should be included in the change in risk estimate
Response
The population of welds that need to be included in the change in risk assessment includes all welds receiving NDE except for those that receive only a surface examination and are not susceptible to outside diameter attack (e.g. ECSCC). This population includes so-called risk category 6 and 7 locations, which are not required to be included in the RI-ISI delta risk assessment.
It is the intent of the Code Case authors to update N-716 to reflect this requirement (i.e.
exclusion of surface only examinations without outside diameter attack) as well as any other relevant feedback from the pilot plant process.
(b)
If all welds that were or will be inspected are included in the change in risk estimates in Table 3.4-1 and 3.4-2 in your submittal, how are the CCDP, CLERP, and the failure frequency estimated for low-safety-significant (LSS) welds?
Deleted: Is this reflected in the N716 application results?¶ Deleted: 14.7
Response
For CCDP/CLERP, values of 1E-4/1E-5 were conservatively used. The rationale for using these values is that the change is risk evaluation process of N-716 is similar to that of the EPRI RI-ISI methodology. As such, the goal is to determine CCDPs/CLERPs threshold values. For example, the threshold values between High and Medium Consequence Categories is 1E-4 (CCDP) / 1E-5 (CLERP) and between Medium and Low Consequence Categories are 1E-6 (CCDP) / 1E-7 (CLERP) from the EPRI RI-ISI Risk Matrix. Using these threshold values streamlines the change in risk evaluation as well as stabilizes the update process. For example, if a CCDP changes from 1E-5 to 3E-5 due to an update, it will still be below the 1E-4 threshold value and the change in risk evaluation would not need to be updated.
The above values were compared to the DC Cook internal flooding study. The CCDPs for in scope LSS Class 2 piping previously being inspected is less than 1E-4 and there were no containment bypass breaks; therefore the 0.1 conditional LERF is also reasonable. The values are consistent with and conservatively above any CCDP value obtained for CNP in-scope Class 2 piping, and that the CLERP value is an appropriately scaled value.
With respect to assigning failure potential for LSS piping, the criteria are defined by Table 3 of the Code Case. That is, those locations identified as susceptible to FAC (or another mechanism and also susceptible to water hammer) are assigned a high failure potential. Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion or stress corrosion cracking are assigned to a medium failure potential and those locations that are identified as not susceptible to degradation are assigned a low failure potential.
In order to streamline the application, a review was conducted to verify that the LSS piping was not susceptible to FAC or water hammer. This review was conducted similar to that done for a traditional RI-ISI application. Thus, the high failure potential category is not applicable to LSS piping. In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g. to determine if thermal fatigue is applicable), these locations were conservatively assigned to the medium failure potential (Assume Medium in Table 3.4-1) for use in the change in risk assessment. Experience with previous RI-ISI applications shows this to be generally very conservative.
(5)
Under Section 3.3 on page 8, your submittal identifies 4 primary guidelines on selecting inspection locations, or 6 guidelines if each sub-bullet in (1) is counted as a guideline. Please describe briefly how each of these six guidelines was applied (e.g., how many inspections were influenced by the guideline and if application of the guideline resulted in changes to the original locations) when you were selecting inspection locations.
Also, were any inspections added due to change in risk considerations?
Response
The process of defining the inspection population of an N-716 application is an iterative process. The first step is to define the scope of HSS welds on a per system basis. As a Deleted: (PRA-FLOOD-014, Attachment A) - CNP needs to review flood groups etc to make sure that the CCDPs greater than 1E-4 are not class 2 AMSE scope etc.¶
starting point, N-716 requires that 10% of the HSS welds, on a per system basis, be selected for inspection (see attached Table 5-1, column entitled HSS). The next step is to assure that 10% of Class 1 welds are selected (see attached Table 5-1, column entitled Class 1). It should be noted that a lesson learned from the CNP application is that this requirement could be more clearly stated in N-716 and it is the authors intent to revise the code case to reflect this and other lessons learned, as applicable. The next step is to assure that 25% of locations identified as potentially susceptible to some type of degradation mechanism be selected (see attached Table 5-1, column entitled DMs).
The next step is to confirm that two thirds of the identified inspections for the RCPB are within the first isolation valve or move inspections from between the two isolation valves to within the first isolation valve to compensate, if necessary (see attached Table 5-1, column entitled RCPBIFIV). The next step is to confirm, or select if necessary, so that 10% of the RCPB that lies outside containment is inspected (see attached Table 5-1, column entitled RCPBOC). Finally, inspections are chosen so that 10% of the break exclusion region (BER) populations are chosen (see attached Table 5-1, column entitled BER). Again, this may have already been accomplished by the preceding criteria, but needs to be confirmed or adjusted accordingly.
Depending upon how the element selection process is ordered, it may be necessary to iterate once or twice to assure the criteria are met. Because of rounding up, the selection being done on a system-by-systems basis, and the multiple criteria, it is expected that a greater than a 10% inspection population will be attained (e.g., CNP witnessed 10.2% for Unit 1 and 10.1% for Unit 2).
With respect to change in risk considerations, no changes to the number or locations of inspection were required.
Deleted: GGNS
Table 5-1 (1)
System Unit Selections HSS Class 1 DMs RCPBIFIV RCPBOC BER Required 67 of 662 67 of 662 8 of 30 45 n/a n/a Unit 1 Actual 67 of 662 67 of 662 16 of 30 67 n/a n/a Required 67 of 669 67 of 669 7 of 26 45 n/a n/a RC Unit 2 Actual 67 of 669 67 of 669 12 of 26 67 n/a n/a Required 7 of 70 7 of 70 8 of 32 (2) 5 n/a n/a Unit 1 Actual 7 of 70 7 of 70 7 of 32 (2) 7 n/a n/a Required 7 of 64 7 of 64 9 of 34 (2) 5 n/a n/a CS Unit 2 Actual 7 of 64 7 of 64 7 of 34 (2) 7 n/a n/a Required 5 of 48 3 of 22 n/a 2
n/a n/a Unit 1 Actual 5 of 48 5 of 22 n/a 2
n/a n/a Required 6 of 55 3 of 27 n/a 2
n/a n/a RH Unit 2 Actual 6 of 55 6 of 27 n/a 2
n/a n/a Required 45 of 442 45 of 442 12 of 47 8 of 32 (3) 1 of 9 n/a Unit 1 Actual 45 of 442 45 of 442 14 of 47 8 of 32 (3) 2 of 9 n/a Required 46 of 457 46 of 457 11 of 43 8 of 32 (3) 1 of 9 n/a SI Unit 2 Actual 46 of 457 46 of 457 12 of 43 8 of 32 (3) 2 of 9 n/a Required 22 of 214 n/a 2 of 8 n/a n/a n/a Unit 1 Actual 22 of 214 n/a 2 of 8 n/a n/a n/a Required 20 of 200 n/a 2 of 8 n/a n/a n/a FW Unit 2 Actual 20 of 200 n/a 2 of 8 n/a n/a n/a (1) For columns entitled HSS, Class 1, DMs, RCPMOC and BER, the information provided is in the format of number of inspections per population of welds (e.g., a 10% requirement for a population of forty (40) welds would be 4 of 40). For the column entitled RCPBIFIV, this criterion is that 2/3 of the Class 1 inspections be inside the first isolation valve. Thus, this column identifies, on a per system basis, how many inspections were required per this criterion (row entitled Required) and how many were actually selected to meet this criterion (row entitled Actual).
(2) Per Section 4(b)(1) of N-716, a minimum of 25% of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected for examination. Per Section 4(b)(2), if the examinations selected per Section 4(b)(1) exceed 10% of the total number of high safety significant welds, the examinations may be reduced by prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that as least 10% of the high safety significant population is inspected. This requirement was applied to the CS system.
(3) A modified element selection approach was implemented for the SI system based on lessons learned to address the requirement that 2/3 of the Class 1 examinations be located between the first isolation valve (i.e., isolation valve closet to the RPV) and the RPV per Section 4(c) of N-716. For CNP, only 32 of 442 Class 1 welds for Unit 1 and 32 of 457 Class 1 welds for Unit 2 are located inside the first isolation valve.
A 25% sampling of the total number of welds located inside the first isolation valve was alternatively selected for examination.
(6)
The relationship between N-716's guideline that "any piping segment whose contribution to core damage frequency (CDF) is greater than 1E-6/year is a high safety significant (HSS) segment," and the EPRI topical guidelines for safety significant categorization is unclear. For example, a low consequence segment in the EPRI Topical methodology has a CCDP less than 1E-6, an identical numerical value but a different metric than the 1E-6/year guideline in N-716. Page 3-8 in the EPRI Topical provides an explanation that the CCDP and conditional large early release probability (CLERP) ranges were selected "to guarantee that all pipe locations ranked in the low consequence category do not have a potential CDF impact higher than 1E-8 per year or a potential large early release frequency (LERF) impact higher than 1E-9 per year." Inspection of Table 3.1-1 and 3.1-2 in your submittal also indicates that there are no entries in the "CDF > 1E-6" column indicating that no segments in the CNP Units 1 and 2 flooding PRA exceeded this guideline.
(a)
The N-716 code case section 2(5) does not include a LERF guideline analogous to the CDF guideline, and Table 3-1 and 3-2 in your submittal includes a column for CDF but not for LERF. Please explain why a LERF guideline is not included as a guideline in parallel with CDF.
Response
N-716 provides five criteria for determining the classification of welds. The CDF guideline for PRA internal flood segments was added to provide additional margin, as applicable, to the initial scope of high safety significant welds (i.e. a belts and suspenders approach). As discussed in the whitepaper, N-716 is based upon lessons learned from a large number of risk-informed applications (e.g. RI-ISI, RI-BER). With respect to defining the scope (e.g. HSS vs. LSS), these insights include both the impact on CDF and LERF (e.g. RI-BER insights). In the whitepaper, eight plants (4 BWRs, 4 PWRs) were compared to the N-716 criteria and N-716 was shown to provide for more inspections than traditional RI-ISI approaches even when the criterion of section 2(5) is not used.
Additionally, as a final step, N-716 requires an assessment of the impact on plant risk which includes both CDF and LERF. This change in risk assessment includes so-called risk category 6 and 7 locations, which are not required to be included in the EPRI RI-ISI delta risk assessment. Risk acceptance criteria for these metrics, are consistent with other RI-ISI applications and meet Reg. Guide 1.174 criteria.
(b)
Please provide a discussion justifying the guideline value for CDF selected in section 2(5) in N-716 (i.e., 1E-6/year).
Response
N-716 provides five criteria for determining the classification of welds. The CDF guideline was added to provide additional margin, as applicable, to the initial scope of high safety significant welds (i.e. a belts and suspenders approach). As discussed in the
whitepaper, N-716 is based upon lessons learned from a large number of risk-informed applications (e.g. RI-ISI, RI-BER). In the whitepaper, eight plants (4 BWRs, 4 PWRs) were compared to the N-716 criteria and N-716 was shown to provide for more inspections than traditional RI-ISI approaches even when the criterion of section 2(5) is not used.
Section 2(5) of N-716 provides an additional criterion that can only potentially increase the scope of high safety significant locations (i.e. will only increase the number of inspections). Consistent with EPRI TR-112657 (Section 3.3.2), the value of 1E-6 (CDF) was chosen as a value that is suitably small and is consistent with the decision criteria for acceptable changes in CDF found in Reg. Guide 1.174. Further, the guideline value is consistent with the philosophy found in EPRI TR-105396 (PSA Applications Guide).
Allocating resources (e.g. NDE) on components below this guideline value (e.g. section 4.2.2) will provide negligible risk benefit while expending unnecessary worker dose and radwaste.
Finally, the assessment of the impact on plant risk (Section 5 of N-716) due to implementing N-716 provides an additional level of assurance that the overall impact on plant risk (CDF and LERF) will be acceptably low. Risk acceptance criteria for these metrics, are consistent with other RI-ISI applications and meet Reg. Guide 1.174 criteria.
This change in risk assessment also includes so-called risk category 6 and 7 locations, which are not required to be included in the EPRI RI-ISI delta risk assessment.
(c)
Please provide a list of all the piping segments that were compared to the >1E-6/year criteria along with the CDF and LERF estimates, the pipe failure frequency, and the CCDP and conditional large early release probability for each segment.
Response
The scope of piping reviewed against this criterion consisted of Class 2 piping not classified as HSS (e.g. BER), Class 3 and non nuclear safety piping. The updated Cook flooding PRA was used to conduct this comparison. The updated Cook flooding PRA was performed consistent with the Draft ASME RA-Sa-2003, Appendix B guideline.
That is, the internal flooding PRA was performed by defining flood zones, identification of flood zone contents (e.g., important equipment), flood zone flood sources and propagation pathways, a qualitative screening analysis, and a quantitative analysis of the remaining potentially important flood scenarios.
The bounding, screening quantitative analyses resulted in all flood zones/groups falling below the 1E-06 CDF criterion except two dominant contributors. The first involved a failure of a fire protection line in the Auxiliary Building which was postulated to flood the Electrical Switchgear Train A DC distribution panel room (CDF contribution of 6.11E-06). The second involved failures of the circulating water system in the Condenser Pit (CDF contribution of 3.75E-06).
Deleted: Cable Enclosure, Battery Room and Battery Charger
Based on the above, more detailed analysis was conducted that reflected a plant modification (fire protection line) and more realistic analyses (e.g. revised HEP) so that these scenarios now fall below the 1E-06 CDF criterion.
With respect to LERF, please see response to 6(a)
(d)
Please provide any observations made during any independent reviews of the CNP flooding PRA or observations from the internal events review that are also applicable to the flooding analysis.
Please describe how these observations have been resolved such that there is confidence that segments that have a CDF greater than the guideline value have been identified.
Response
There was only one Internal Flooding related Facts and Observations (F&O) from the Cook PRA Peer Review process. That F&O was as follows:
Flood barriers were not treated probabilistically. All flood barriers were assumed to function. Back flow through drains was also not assumed to occur.
The flooding analysis screened away all rooms except the turbine building basement. The screening criteria considered pipe spray mode only (i.e., no ruptures), which resulting in the screening out of all rooms.
This is level A significance, since the flooding CDF is very low (2E-7), based on screening away of all rooms using erroneous criteria.
That single F&O was resolved by generation of an updated Cook flooding PRA. Which, as noted in the response to Question 6c above, was performed consistent with the then Draft ASME RA-Sa-2003, Appendix B flooding study guidelines. That is, the internal flooding study was performed by defining flood zones, identification of flood zone contents (e.g., important equipment), flood zone flood sources and propagation pathways, a qualitative screening analysis, and a quantitative analysis of the remaining potentially important flood scenarios.
(e)
Page 3 of your submittal states that '[i]nternal flooding was recently addressed (2006) to complete the effort to address all Westinghouse Owners Group certification Level A and B F&Os." To the extent not discussed in the response to RAI 6(d), please explain what "addressed" means. Were changes made to a flooding analysis? If changes were not made, how are the F&Os addressed? How Formatted: Normal, Space After: 0 pt Deleted: CNP PRA to respond
were the changes that were made, or the explanation for not requiring changes, reviewed for technical adequacy?
Response
Addressed", as used in the submittal, means that the Cook internal flooding analysis was updated to meet the intent of the flooding F&O, as noted above in 6(d). The prior Cook flooding analysis was the analysis from the original Cook IPE flooding evaluation. The flooding update was accomplished over the 2005-2006 time frame, using then Draft Addenda B to ASME RA-Sa-2003 technique & criteria for determining flooding susceptibility, screening areas/systems from consideration, etcetera. This update was captured in the form of calculations, in accordance with plant procedures for non-safety related calculations. These calculations, while not safety related, underwent an in-house, independent review (e.g., calculation preparer and reviewer) process. The update resulted in substantial changes in the flooding evaluation from the initial IPE evaluation upon which the flooding F&O was based. The updated flooding evaluation relied on the 2005 internal events PRA model to address various flooding scenarios that were not screened out of consideration.
This updated PRA flooding analysis initially found a scenario that produced a CDF in excess of 1E-6. The scenario involves a fire hose station mounted in a small room housing the Train A DC distribution panels, that also had the potential to impact the Train B DC distribution system. A minor plant hardware modification to seal the associated Train B DC panel subsequently reduced the impact of this scenario, removing it from the list of scenarios requiring further detailed analysis. The Circulating Water System was involved failures of the circulating water system in the Condenser Pit (CDF contribution of 3.75E-06).
(7)
Page 12 describes how the CCDP and CLERP of different types of HSS pipe breaks are estimated in support of the change in risk estimates. Some values appear to be derived from representative sequences from the PRA models while others are directly estimated.
For example, bounding values for pipe breaks that result in isolable loss-of-coolant accidents (LOCAs) are directly estimated as the product of the CCDP from unisolable LOCAs and the probability of a motor operated valve failing to close on demand. Direct estimation can be very analyst-specific and essentially bypasses the PRA peer review process upon which the NRC relies to minimize the staff review of the plant-specific PRA for each risk-informed submittal.
(a)
Please identify events modeled in the CNP PRA that are similar to the directly estimated values on page 12 of your submittal or further clarify why the your PRA can not be used to develop the required estimates (these appear to be ILOCA, PLOCA, PILOCA-OC, and PILOCA-IC).
Formatted: Normal, Space After: 0 pt, Don't keep with next Formatted: Font: (Default) Times New Roman, Not Highlight Formatted: Font color: Black Deleted: CNP PRA to respond.
If applicable events in the PRA can be identified, please provide a description of these events and the bounding CCDP and CLERP values for these types of breaks derived from the PRA.
Response
The CNP PRA does not explicitly model potential and isolable LOCA events, because the LOCA initiators in the PRA do not distinguish break location. The N-716 methodology must evaluate these segments individually, thus it is necessary to estimate their contribution. This is estimated by taking the LOCA CCDP and multiplying this by the valve failure probability.
(b)
In the Table on page 12, please describe the difference between row 2, isolable LOCA (assumed to be inside containment), and row 5, potentially isolable LOCA inside containment. In what category would a pipe break that relied on a MOV that does not close automatically but that could be closed remotely by a manual action be placed?
Response
The isolable LOCA (row 2) is a segment downstream of an AOV that automatically isolates on low pressurizer level (template table on page 12 has typo indicating MOV rather than AOV). MOVs have a failure probability that is slightly larger than AOVs, which therefore provides a slightly higher CCDP. For conservatism, the high consequence rank is maintained. The potential LOCA (row 5) is a segment downstream of a normally closed valve, in this case a check valve. Thus, the CCDP is estimated as the product of check valve rupture and LOCA CCDP. This particular segment also includes some piping downstream of a MOV that does not get an automatic signal, thus credit for another isolation valve was not taken. Since there is uncertainty with regard to the operators ability to detect this break location in time to prevent a LOCA, operator action was not credited.
(c)
Row 6, "Class 2 SDC - IC" states that the CCDP and CLERP are "[e]stimated based on a loss of shutdown cooling during mid-loop operation." Are these values intended to develop the safety significance of these segments during shutdown, or as surrogates for power operation. If these values are intended as surrogates for power operation, please explain why these values are reasonable surrogates. If not intended as surrogates for power operation, how was the safety significance of these segments during power operation addressed.
Response
Since there are 2 normally closed valves during power operation, the CCDP for power operation is clearly <1E-4 as summarized below (this result is consistent with a number of RI-ISI applications):
- The potential LOCA scenarios require 2 valves in series to fail open and then this would be multiplied by LOCA CCDP
- The injection paths could also fail during an accident demand, but there are redundant backup injection paths and the CCDP for this event required the probability of challenge times the CCDP for the backup paths.
As a result, it was assumed that pipe break during shutdown operation could be more important and it was assumed to have a 1E-4 CCDP based on qualitative reviews on several previous RI-ISI applications. The reference to mid-loop could be deleted as it could be misleading in that the table was meant to provide a general reference to shutdown configurations, not just mid-loop.
(d)
The last row in the Table on page 12 includes an entry labeled "Class 2 LSS". What characteristics results in a "Class 2 LSS" designation? The same entry further states that the CCDPs and CLERPs of pipe ruptures associated with these welds are
"[e]stimated based on upper bound for Medium Consequence." Please provide a discussion explaining why selecting these values is appropriate.
Response
The Class 2 LSS designation is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criterions of section 2(a) of N-716 (e.g. not part of the BER scope). With respect to CCDPs/CLERPs, please see the answer to question 4(b).
(e)
The ASME standard RA-Sa-2003, element IE-C12 discusses the evaluation of the likelihood of a interfacing system LOCA. In which category does the interfacing system LOCA belong in your Table on page 12?
Response
The PILOCA-OC break location on page 12 applies to the category breaks in piping connected to RCPB outside containment. For CNP, a 1.0 CCDP=CLERP was used for piping outside containment and connected to the RCPB (the CCDP only credited valve failures required to cause the LOCA outside containment). A conservative estimate of CCDP can be used for this application as long as it supports the determination that delta risk is low. More realistic calculations would only be required if these simplified approaches indicated potentially unacceptable risk increases.
(8)
Under Section 2.2 on Page 4, you state that, "[t]he requirements of MRP-139 will be used for inspection and management of primary water stress corrosion cracking (PWSCC) susceptible welds and will supplement the RIS_B Program selection process." Please describe what is meant by "supplement." How will the PWSCC degradation mechanism be addressed, as any other mechanism or differently? How will any inspections that might be required by MRP-139 be credited in the RIS_B program?
Response
MRP-139 inspection schedule will be followed. All of the pressurizer nozzle butt welds have had weld overlays installed. The inspection schedule will be in accordance with Relief Requests ISIR-15, 20 & 21. This requires 25% of the overlays to be inspected during the interval. The remaining butt welds are in the Unit 1 Reactor vessel nozzles.
The reactor vessel nozzles will be inspected and/or mitigated using the guidance in MRP-139.
(9) Is the guideline to examine a minimum 10% of all HSS welds, or 10% of all HSS butt welds, or 10% of all HSS butt welds >= 4 NPS?
Response
Please see response to Question 3(a)
(10)
Under Section 3.4 on Page 11, your submittal states "the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements." Consistent with this, the total change in risk in the two tables on page 14 is always negative. Is Cook committing to ensuring that these total risk numbers will be maintained at or below 0 as it monitors the program over time as described in Section 4 of your submittal?
Response
The change in risk will meet the acceptance criteria per section 5 (d) of Code Case N-716. This is consistent with the acceptance criteria in EPRI TR-112657.
Deleted: ASME Section XI, Appendix Q
PART 2 - Second Set of Questions By a previous e-mail (Accession No. ML070890463) I sent you draft RAI questions from the PRA branch. Following are draft RAI questions from our Piping and NDE Branch (Andrea Keim, reviewer). We request that you discuss these two sets of draft RAI questions and your proposed response with us either in a conference call, or in a public meeting:
(1)
Footnote 2 for the table on page 9 of the licensees submittal indicates that 240 Class 2 welds are HSS, yet only 22 welds are selected for inspection at Unit 1 and 228 Class 2 welds are HSS yet only 20 welds are selected for inspection at unit 2. These selections do not appear to meet the 10% requirement for HSS locations. Please explain this discrepancy.
Response
Per Section 4 of Code Case N-716, ten percent of the high safety significant welds shall be selected for examination. Subparagraphs 4(a) through 4(f) of Code Case N-716 specify how the 10% sampling shall be distributed. These requirements are addressed in the response to Question 5 from the first set of RAIs. Code Case N-716 does not require that a 10% sampling of the ASME Code Class 2 welds designated as HSS be selected for examination. As stated in the response to Question 5 from the first set of RAIs, N-716 will be revised to explicitly state that a 10% selection of Class 1 welds is required, but this same requirement does not apply to Class 2. This selection philosophy, as it pertains to Class 1 and 2 piping welds, is identical to that implemented in EPRI TR-112657 that has been approved by the NRC.
(2)
Section 5 of the licensees submittal states that the licensee will implement the RIS B program during the plants third period of the current (third) inspection interval by performing 66% of the inspection locations selected for examination per the RIS B process for each unit. Describe how the licensee will determine which examinations to perform during the remainder of the third 10-year ISI interval.
Response
Prior to developing the RIS_B Program, CNP had planned to inspect locations scheduled for examination in the traditional ASME Section XI inspection program. Examination activities during refueling outages are planned far in advance. In general, only designated plant areas and components are accessible for examination during a given refueling outage due to other ongoing plant maintenance and modification activities. Hence, any location previously scheduled for examination in the third period via the traditional program will remain scheduled for examination in the third period if the location has also been selected for RIS_B Program purposes. To complete the sample size, additional locations will be selected, if necessary, to Formatted: Normal, Indent: Left:
0", Space After: 0 pt Deleted: Some of these inspections have been completed in the last unit outages. The remaining scope will be divided over the next series of outages.
The next two Unit 1 and the next two Unit 2 outages have the required inspections scheduled.¶ Grand Gulf Response for Consideration by CNP)¶
achieve equal representation of the degradation mechanisms. Other factors such as accessibility and scaffolding requirements will also be factored into the selection process.
(3)
Please describe how volumetric examinations will be performed. At a minimum, will volumetric examinations include the volume required for ASME Section XI examinations? Will ASME Section XI, Appendix VIII qualified examiners and procedures be used for all volumetric exams? Will the examination volume be scanned for both axial and transverse indications for all exams? Please describe and justify your answers.
Response
Volumetric examinations will be performed as required by Table 1 of N-716. The table requires an examination volume as defined in the ASME Section XI IWB figures. This would require examination of at least the ASME Section XI volume. (More volume may be required based on the notes on Table 1.) N-716 does not take any exceptions to the paragraphs of the Code that govern volumetric examinations and the request for alternative does not take exception to any 10 CFR limitations. Therefore, I&M will examine these welds using the same personnel and procedure requirements as a traditional Section XI piping volumetric examination (4)
Please describe how preservice examinations will be performed for repair/replacement activities. Include what repair/replacement items will receive preservice examination.
Response
For preservice examinations, I&M will follow the rules contained in Section 3.0 of N-716. Welds classified HSS require preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1. Welds classified as LSS do not require preservice inspection.
(5)
Page 10 discusses additional examinations. Please describe what will be used to perform the engineering evaluation to determine the cause of any unacceptable flaw or relevant condition. Recent industry practice has been to perform corrective actions (i.e.,
overlays, replacement, etc.) prior to a root cause being determined (e.g., use of a qualified procedure and personnel).
Response
Any unacceptable flaw will be evaluated per the requirements of ASME Code Section XI, IWB-3500 and/or IWB-3600. As part of performing evaluation to IWB-3600, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation. If the flaw is found unacceptable for continued operation, it will be Deleted: ¶ Volumetric examinations are performed using UT or RT as outlined in the program. All examinations will be performed using ASME B&PV Code Section XI, Appendix VIII qualified procedures, equipment, and examiners.
The axial and transverse scanning will be in accordance with the qualified procedures.¶ Alternate Response (Grand Gulf)¶ Deleted: Entergy Deleted: Pre-service examinations will be performed per the ASME Code Section XI requirements. These examinations will provide a baseline data for future use to determine service related degradation.¶ Alternate Response (Grand Gulf)¶ Deleted: Entergy Deleted: The ASME Boiler & Pressure Vessel Code,Section XI, IWB-3500 and IWB-3600 methodology will be used to evaluate flaws. If the flaw is unacceptable, it will be repaired in accordance with IWB-4000 and the applicable Code Cases. The industry practice such as performing preemptive overlays, is dictated by practical considerations. The single most governing concern in older plants is that some locations have joint designs that do not permit Performance Demonstration Initiative examinations to be performed.
The result is a preemptive weld overlay.
The resultant joint can be inspected in compliance with Appendix VIII and meets the full structural capability.¶ Alternate Response (NGC)¶
repaired in accordance with IWB-4000 and/or applicable ASME Section XI Code Cases.
The need for extensive root cause analysis beyond that required for IWB-3600 evaluation will be dependent on practical considerations (such as the practicality of performing additional NDE or removal of the flaw for further evaluation during the outage)
(a)
In some cases no materials are removed for metallurgical analysis. Please discuss the process used for this engineering evaluation, how will it be documented, and will the Nuclear Regulatory Commission be involved in the process?
Response
The process for ordinary flaws is to perform the evaluation using the ASME B&PV Code Section XI. If the flaw meets the criteria, then it is noted and the appropriate successive examinations are scheduled.
The Nuclear Regulatory Commission is involved in the process at several points. For preemptive overlays, a Relief Request is usually needed for the design and installation.
Should the flaw be discovered during the examination, a notification per 10CFR 10.72 or 10CFR10.73 may be made. IWB-3600 requires the evaluation to be submitted to NRC.
Finally, an NIS-1 and NIS-2 forms are submitted which summarizes the inspections and repairs performed during the outage.
(b)
Discuss what process will be used to perform fracture mechanics evaluations
Response
ASME B&PV Code Section XI, IWB-3600 provides the rules for flaw evaluation and fracture mechanics. The results of the evaluation are required to be submitted to the NRC.
(c)
Discuss under what conditions would there be no additional examinations.
Discuss how the licensee will document its justification.
Response
If the flaw is original construction or otherwise acceptable, Code rules do not require any additional inspections. If the nature and type of the flaw is service induced, then similar systems or trains will be examined. The documentation requirements will be documented in the Corrective Action Program and a summary will be submitted in the NIS-1 package.
(6)
Page 10, Section 3.3.2 Program Relief Requests, provides guidance. For program relief requests the licensee refers to the process outlined in Reference 2. Recently there Formatted Deleted: Primary Water Stress Corrosion Cracking of Alloy 600/82/182 butt weld locations are limited to the Unit 1 hot and cold legs. All of the other Class 1 butt welds have had overlays applied.
Deleted: ¶ Deleted: nd Deleted: Alternate Response (Grand Gulf)¶ Based on the NRCs questions and recent industry events, Entergy agrees that the process contained in GG-ISI-002 could be difficult to codify as written and difficult to consistently perform and, thus, difficult to regulate. Therefore, Entergy will revise the submittal to use the additional examination criteria contain in N-716, Section 6. These requirements are concise and can be implemented consistently each occurrence.
Additionally, these requirements are the same as those that ASME have recently approved in Appendix R¶ Deleted: Alternate Response (Grand Gulf)¶ See the response to Question 5a¶ Deleted: Alternate Response (Grand Gulf)¶ See the response to Question 5a¶
have been problems associated with giving relief limited examinations from risk-informed ISI program items. For limited examinations of RIS_B selected items please describe the process for assessing limited examination coverage. Discuss whether additional examinations will be performed, and whether additional techniques will be used to improve examination coverage. Discuss how the effect on risk of the incomplete examination coverage will be assessed. In what time frame will relief requests be submitted?
Response
Consistent with previously approved RI-ISI submittals (e.g. ANO, Unit 2 SER), I&M will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section XI examinations. Experience has shown this process to be weld-specific (e.g., joint configuration). As such, the effect on risk, if any, will not be known until that time. Relief requests will be submitted per the guidance of 10 CFR 50.55a(g)(5)(iv) within one (1) year after the end of the interval.
(7)
Page 10 also discusses that Relief Requests ISIR-005 and ISIR-006 will be withdrawn. Please discuss why these requests will be withdrawn. Also the licensee states that pipe-to-flue head welds in the feedwater system are included in the scope that is designated high safety significant, yet have not been selected for examination. Describe why none of these welds are selected for examination.
Response
During the development of the risk-informed template process, the NRC requested that licensees address the impact of the risk-informed application on existing plant ISI Program relief requests. The NRC requested notification in the template submittal of any relief requests that would be modified or withdrawn as a result of the change in inspection philosophy. For the DC Cook N-716 application, this impact is addressed in Section 3.3.2 of the plant template submittal. Further explanation is provided below.
Feedwater Pipe to Flued Head Welds These locations are included in the system boundaries (i.e., steam generator to the outer containment isolation valve) designated high safety significant, but were not selected for RIS_B examination. I&M did not choose these locations as part of the 10% HSS examination sampling required for the feedwater system because they are inaccessible and because no degradation mechanisms were identified. In addition, it should be noted that these locations are not mandatory selections per the 1989 Edition of ASME Section XI which is the Code of record at DC Cook for the current 3rd interval ISI Program. As such, a relief request is not required.
Main Steam Pipe to Flued Head Welds The main steam system in its entirety is designated low safety significant and is therefore not subject to RIS_B examination.
Similar to above, it should be noted that these locations are not mandatory selections per Deleted: No additional examinations will be performed, but other locations on the same system may be selected if they are subject to the same temperature, pressure, and potential degradation mechanisms, save dissimilar metal welds. ¶ The Limited Examination Coverage inspections have minimal risk change since the root of the weld can be seen, albeit typically from only one direction.
POD may. ¶
¶ Relief Requests for Limited Coverage are typically submitted at the end¶ of the Interval as required by 10CFR50.55a.¶ Alternate Deleted: for Grand Gulf¶ Deleted: Entergy
the 1989 Edition of ASME Section XI which is the Code of record at DC Cook for the current 3rd interval ISI Program. As such, a relief request is not required.
(8)
Section 3.3.2 states that an attempt was made to select locations for examination such that a minimum >90% coverage is attained. Discuss how this attempt was conducted.
If less than 90% examination is completed, discuss whether additional weld(s) will be examined to compensate for the limited examination coverage.
Response
As discussed in EPRI TR-112657, accessibility is an important consideration in the element selection process of a RI-ISI application. As such, for the CNP N-716 application, locations have generally been selected for examination where the desired coverage is achievable. This is typically accomplished by utilizing previous inspection history, plant access considerations, and knowledgeable plant personnel. However, some limitations will not be known until the examination is performed since some locations will be examined for the first time.
In addition, other considerations may take precedence and dictate the selection of locations where greater than 90% examination coverage is physically impossible. This is especially true for element selections where a degradation mechanism may be operative (e.g., risk categories 1, 2, 3 and 5 of EPRI TR-112657). For these locations, elements are generally selected for examination on the basis of predicted degradation severity. For example, in the emergency core cooling system (ECCS) injection lines of PWRs, the piping section immediately upstream of the first isolation check valve is considered susceptible to intergranular stress corrosion cracking (IGSCC), assuming a sufficiently high temperature and oxygenated water supply. The piping element (pipe-to-valve weld) located nearest the heat source will be subjected to the highest temperature (conduction heating). As such, this location will generally be selected for examination since it is considered more susceptible than locations further removed from the heat source, even though a pipe-to-valve weld is inherently more difficult to examine and obtain full coverage than most other configurations (e.g., pipe-to-elbow weld). In this example, less than 90% coverage of this location will yield far more valuable information than 100%
coverage of a less susceptible location.
For locations with no identified degradation mechanisms (i.e., similar to risk category 4 of EPRI TR-112657), a greater degree of flexibility exists in choosing inspection locations. As such, if at the time of examination an N-716 element selection is found to be obstructed, a more suitable location may be substituted instead.
Therefore, I&M will review each instance of limited coverage and take the appropriate steps (e.g., relief requests) consistent with its impact on the basis of the N-716 application.