2CAN020804, Responses to Request for Additional Information Request for Alternative ANO2-R&R-004, Revision 1
| ML080520186 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/20/2008 |
| From: | James D Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2CAN020804, TAC MD5250 | |
| Download: ML080520186 (24) | |
Text
2CAN020804 February 20, 2008 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Responses to Request for Additional Information Request for Alternative ANO2-R&R-004, Revision 1 Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6
REFERENCES:
- 1. Entergy Letter to NRC dated April 17, 2007, Request for Alternative ANO2-R&R-004, Revision 1 Request to Use Risk-Informed Safety Classification and Treatment for Repair / Replacement Activities in Class 2 and 3 Moderate Energy Systems (CNRO-2007-00015)
- 2. Entergy Letter to NRC dated August 6, 2007, Request for Alternative ANO2-R&R-004, Revision 1 Response to NRC Request for Additional Information (CNRO-2007-00028)
- 3. NRC Letter to Entergy dated December 20, 2007, Request for Additional Information re: ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair /
Replacement Activities in Class 2 and 3 Moderate Energy Systems (TAC NO. MD5250)
Dear Sir or Madam:
By letter dated April 17, 2007 (Reference 1), Entergy Operations, Inc. (Entergy) proposed to use a risk-informed safety classification and treatment for repair / replacement activities in Class 2 and 3 moderate energy systems at Arkansas Nuclear One, Unit 2 (ANO-2). In response to the NRC, Entergy provided a test case, an example, that demonstrated the process (Reference 2).
Several discussions with the NRC have taken place since transmittal of the aforementioned letter and a meeting at the ANO-2 site in November 2007, resulting in a Request for Additional Information (RAI) by the NRC Staff (Reference 3). Entergys response to the RAI is included in Enclosure 1.
This letter contains one new commitment. This commitment is identified in Enclosure 2.
Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4619 Dale E. James Manager, Licensing Arkansas Nuclear One
2CAN020804 Page 2 of 3 If you have any questions or require additional information, please contact Bob Clark at 479 858 4663.
Sincerely, DEJ/rwc
Enclosures:
- 1.
Response to Request for Additional Information Relating to Request for Alternative ANO2-R&R-004, Revision 1
- 2.
List of Regulatory Commitments cc:
Mr. J. S. Forbes (ECH)
Mr. T. G. Mitchell (ANO)
Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 U.S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS O-7 D1 Washington, DC 20555-0001
2CAN020804 Page 3 of 3 bcc:
Mr. W. B. Abraham (G-ADM2-LIC)
Mr. C. A. Bottemiller (G-ADM2-LIC)
Ms. S. T. Fontenot (W-GSB-318)
Mr. J. G. Weicks (M-ECH-36)
Mr. K. W. Hall (M-ECH-36)
Mr. D. E. James (N-GSB-64)
Mr. W. J. James (N-GSB-59)
Mr. R. J. King (R-GSB-42)
Mr. J. A. Kowaleski (N-ADM-22)
Mr. M. A. Krupa (G-ADM1-NSR)
Mr. R. S. Lewis (M-ECH-36)
Mr. D. N. Lorfing (R-GSB-42)
Ms. K. A. Maher (R-GSB-42)
Mr. W. D. Sims (N-GSB-60)
Ms. D. S. Waldron (N-GSB-64)
ENCLOSURE 1 2CAN020804 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO REQUEST FOR ALTERNATIVE ANO2-R&R-004, REVISION 1 to 2CAN020804 Page 1 of 12 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO REQUEST FOR ALTERNATIVE ANO2-R&R-004, REVISION 1 NRC Question #1 The example analysis (Example) provided as the enclosure to the August 6, 2007, (Reference 12) letter included a number of pages from ANO-2s October 8, 1998, submittal on risk-informed inservice inspection (RI-ISI). Inspection of the October 8, 1998, submittal indicates that a screening process was applied to some piping. Neither this piping nor the screening processes are discussed in the Example. Instead, the Example only illustrates the comprehensive categorization methodology described in Attachment 1 to the April 17, 2007, submittal (Enclosure) (Reference 11). This process includes a detailed consequence determination for each segment using the appropriate categorization Tables in the Enclosure (e.g., segment CSS-C-01 in the Example). The screening process is apparently applied to some population of piping before, or in lieu of, the detailed analysis. The screening process is not described in the Enclosure or the Example submittals.
Please describe the screening process, including all the guidelines to screen out piping and how these guidelines are appropriate for changing repair and replacement (R&R) activities.
Please describe how the two processes are applied together, beginning with a definition of the initial population of piping and ending with all piping being categorized based on one or more of the processes.
Please identify where the screening process is permitted and/or described in the categorization methodology in the Enclosure.
Response
The above NRC discussion accurately describes the process used for defining the scope of the ANO-2 RI-ISI pilot application. At the time of the ANO-2 RI-ISI pilot study there was ongoing discussion between the industry and the NRC as to the required scope of a RI-ISI application.
That is, would a partial scope RI-ISI application in lieu of a full scope RI-ISI application be acceptable. Since that time, partial scope RI-ISI applications have been found acceptable by the NRC. In fact, the first NRC approved RI-ISI application was Vermont Yankee, which was a Code Case N560 application (i.e. a subset of Class 1 piping).
As a minimum, the ANO-2 RI-ISI program covers that portion of systems subjected to non-destructive examination (NDE) per the deterministic ASME Section XI scope definitions. This will also be true for the RI-RRM application. Any Class 2 or 3 system or portion of systems that are not subjected to the analysis requirements of the proposed methodology will be classified as high safety significant (HSS). That is, any system or portions of systems that used the RI-ISI screening process or was not evaluated in detail per the RI-ISI consequence assessment process will be classified HSS unless they are subsequently evaluated per the proposed RI-RRM methodology.
to 2CAN020804 Page 2 of 12 NRC Question #2 The Example (Reference 12) included a categorization of piping in the containment spray system (CSS) performed in support of the RI-ISI program. The CSS was one of many systems evaluated to support ANO2s RI-ISI program submitted in 1998 and was performed consistent with a draft version of EPRI topical TR-112657 (EPRI Topical). The August 6, 2007, supplemental letter augments the Example (Reference 12) with a discussion of 10 conditions (in the form of questions) that are part of the categorization process in the Enclosure (Reference 11) but that are not included in the EPRI Topical and therefore not in the Example (Attachment 12). No other changes or additional steps were described that would be necessary to convert the RI-ISI categorization into a categorization that satisfies all the guidelines in the Enclosure (Attachment 11). There appears to be, however, some differences between categorization described in the Enclosure and the RI-ISI categorization that are not addressed in the submittals.
In the Enclosure (Reference 11), the first paragraph under, III. Requested authorization, concludes by stating that, the process shall be applied on a system basis, including pressure retaining items and their associated supports. The provided Example (Reference 12) does not include associated supports and (as noted in RAI 1) does not include all piping in the system. Please clarify when, and with what guidance documents, piping segments and associated supports will be categorized before changes to the R&R activities are implemented for any given pipe segment.
The Enclosure (Reference 11) states that the consequence of each pipe segment failure should be evaluated with and without operator action and that the highest category should be chosen. The EPRI Topical used to perform the categorization in the Example (Reference 12) does not require that the highest category be chosen. Please confirm that the analyses done in support of R&R activities (both in the Example (Reference 12) and subsequent analyses) will be consistent with the method described in the Enclosure (Reference 11).
Section I-3.0.1 in the Enclosure (Reference 11) describes the features that shall be provided if credit is taken for operator actions. The last feature is that Operators are trained on the procedures. The EPRI Topical used to perform the categorization in the Example does not require this feature before credit can be taken for operator actions. Please confirm that the analyses done in support of R&R activities (both in the Example (Reference 12) and subsequent analyses) will be consistent with the method described in the Enclosure (Reference 11).
Section I-3.2.2 in the Enclosure (Reference 11) states that 10 questions shall be answered for each medium and low consequence segment. In contrast, the August 6, 2007, submittal (Reference 12) answers each question once for all segments in the system, not for each segment. This may be acceptable when the same answer is applicable to every segment in the system but this may not always be the case. Please confirm that the questions were considered for each segment in the Example, and that the analysis done in support of R&R activities will be consistent with the method described in the Enclosure (Reference 11).
to 2CAN020804 Page 3 of 12
Response
Each of the above bullets is responded to as follows:
First Bullet - Please see the response to question #1 with respect to the scope of the system analyses and classification results. As to supports, Section I-3.2.2(d) of the proposed methodology will be followed. That is, section I-3.2.2(d) requires that A component support, hanger, or snubber shall have the same classification as the highest-ranked piping segment within the piping analytical model in which the support is included.
Second Bullet - Although it could probably be more clearly stated, the EPRI Topical Report (EPRI TR-112657, section 3.3.3.2.2) and the proposed methodology are identical in that where possible, operator actions may be credited and if they are, both cases (i.e. successful and unsuccessful) should be analyzed and the highest consequence ranking used in the final classification. Section I-3.0.1 of the proposed methodology specifically states that The scenario that results in the highest consequence ranking shall be used.1 Entergy is committed to following this process for crediting operator action.
Third Bullet - Again the EPRI Topical could be clearer on this requirement, however, Entergy is committed to following the requirement in the proposed methodology for crediting operator action.
Fourth Bullet - For the example system provided, the questions were considered for each segment in the example system. Analysis done for future systems will be consistent with the proposed methodology described in the Attachment.
NRC Question #3 The NRC has endorsed ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities. The licensees Enclosure (Reference 11) proposes a method similar to Code Case N-660 but with some modifications. Please identify the differences between the proposed method and Code Case N-660, and the basis for acceptability for the differences.
Response
The requested information, minus editorial type changes, is provided in Attachment 1.
NRC Question #4 Sections I-3.1.2 (e) and (f) in the Enclosure (Reference 11) describe how shutdown operations and external events shall be evaluated. The Example (Reference 12) did not address shutdown and external events. Please provide an analysis that includes shutdown and external events in the categorization process consistent with the guidance in the Enclosure (Reference 11).
Response
Consistent with the proposed methodology these events were considered as follows:
to 2CAN020804 Page 4 of 12 Shutdown Operation The consequence evaluation is an assessment assuming the plant is at-power. Generally, the at-power plant configuration is considered to present the greatest risk for piping failures since the plant requires immediate response to satisfy reactivity control, heat removal, and inventory control. By satisfying these safety functions, the plant will be shut down and maintained in a stable state. At-power, the plant is critical, and is at higher pressure and temperature in comparison to shutdown operation. The following assessment was conducted to gain confidence that the consequence ranking during shutdown would not be more limiting.
Pipe segments that are already ranked as HIGH consequence from the evaluation at-power need not be evaluated for shutdown. Those that are already MEDIUM require some confidence that HIGH would not occur due to shutdown configurations. However, the LOW consequences for power operation require more confidence that a HIGH would not occur and some confidence that a MEDIUM consequence would not occur. Recognizing this, a review and comparison of system consequence results for power operation versus potential consequence during shutdown operation was conducted.
The results of the comparison indicate that during at-power operation, for the example system (e.g. CSS), segments are ranked as HIGH, MEDIUM and LOW, depending on the pipe size and resulting impact. During shutdown operation, the frequency of challenging the CSS is reduced significantly. The consequence ranking of CSS segments during shutdown operation would be no more risk-significant than the ranking identified during at-power operation. The at-power ranking is therefore assumed to be bounding.
External Events The potential importance of piping failures during an external event is also considered. The ANO-2 IPEEE was reviewed to determine whether external initiating events, with their potential common cause impacts on mitigating systems, could affect consequence ranking. This information, along with information from other external event PRAs, is considered to derive insights and confidence that consequence ranking is not more significant during an external event. The following summarizes the review for each of the major hazards (seismic, fire and others).
Seismic Challenges - The potential effects of seismic initiating events on consequence ranking is assessed by considering the frequency of challenging plant mitigating systems and the potential impact on the existing consequence ranking. The following summarizes this assessment:
Generally, the CSS piping considered in this evaluation has a seismic fragility capacity much greater than the 0.3g screening value and is not considered likely to fail during a seismic event.
With regard to the impact on mitigating systems, the most likely scenario is seismic induced loss of offsite power. Based on a typical fragility of 0.1g (Reference 9.19 of the CSS evaluation) for loss of offsite power and the seismic hazards developed for the ANO site (References 9.20 and 9.21 of the CSS evaluation), the frequency of seismic induced loss of offsite power is less than 1.0E-4 per year.
to 2CAN020804 Page 5 of 12 Considering the scenario where an induced LOCA and non-seismic failure of one diesel generator occur, there is one train of CSS and one train of the Containment Cooling System (CCS) available. In response to the event, the available CSS train is assumed to fail when demanded with the CCS train providing backup. Assuming probabilities of 0.1 and 1.0E-2 for failure of the diesel generator and the backup train, respectively, and an all year exposure time, the Conditional Core Damage Probability (CCDP) for this scenario is less than 1.0E-06.
Thus, the resulting consequence is LOW.
For the scenario where an induced LOCA occurs and both diesel generators are available, both CSS trains are initially available. However, in response to the event, one CSS train is assumed to fail with the remaining CSS train and both CCS trains providing backup. Assuming a probability of 1.0E-2 for a backup train and an all year exposure time, the CCDP is less than 1.0E-6. Thus, the resulting consequence is LOW. Other seismic scenarios that include challenges to CSS would result in lower CCDPs because additional failures (i.e., failure of PSV to reclose or failure of EFW) must occur. However, the consequence ranking would also be LOW.
Based on the above, the consequence ranking for CSS during a seismic event is enveloped by the at-power consequence ranking.
Fire Challenges - The ANO-2 IPEEE indicates that the fire core damage frequency is dominated by fires initiated outside the containment. Similar to a seismic event, fires do not impact reactivity control or cause a LOCA unless it is due to a stuck open primary safety valve.
Based on fire core damage frequency of 3.5E-05 per year in the ANO-2 IPEEE, a fire induced loss of offsite power frequency is assumed to be less than 1.0E-02 per year.
With regard to the impact on mitigating systems, the most likely scenario fire is induced loss of offsite power. Considering the scenario where fire induced loss of offsite power occurs and both emergency diesel generator trains are available, a transient induced LOCA (i.e., opening and failure of a PSV to reclose) or failure of EFW would challenge the CSS. Reclosure of the PSVs or EFW operation and the intact CSS and CCS trains (assuming one CSS train fails on demand) would provide three or more backup trains for mitigation. Combining the frequency of loss of offsite power with three available backup trains results in a LOW consequence.
Considering the same scenario with loss of one emergency diesel generator would result in two backup trains for mitigation. Hence, the resulting consequence would also be LOW. Since the at-power consequence ranking is already HIGH, MEDIUM or LOW, the resulting consequences during a fire would not be of greater risk significance.
Other External Challenges - Other hazards were screened in the ANO-2 IPEEE and are assumed to have little or no risk significant impact on CSS.
to 2CAN020804 Page 6 of 12 NRC Question #5 Many of the tables referenced in the Example refer to tables in previous documents. Please provide a cross reference relating the tables referenced in the Example to those in the Enclosure.
Response
The requested cross reference is provided as follows:
Location in the Example Wording used in Example (Ref. 12)
Cross Reference to Proposed Methodology (Ref. 11)
Section II. Application, Table 1 - CSS Consequence Assessment Summary, 11th Column entitled Table Used 2-2 Table I-2 Section V. Consequence Information Reports Table 2-2 Table I-2 Section V. Consequence Information Reports Table 2-4 Table I-4 NRC Question #6 The evaluations in the Example always assumed a large pipe break consistent with the guidance in the EPRI Topical. Code Case N-660 requires the assumption of a large pipe break unless leak-before-break can be justified in accordance with NUREG-1061 or the plant configuration precludes the possibility of a large break. The method proposed in your Enclosure (Reference 11) permits assuming a small leak that is determined by analytic evaluations.
Please justify the use of a small break instead of a full break and provide illustrative examples of these analytic evaluations
Response
It is true the evaluations in the Example (Reference 12) as well as the other ANO-2 systems included in the RI-ISI scope assume a large break. The concept and technical basis for assuming something less than a large break is the subject of current industry / NRC interactions. As such, Entergy proposes for purposes of this relief request to continue to use the RI-ISI large break assumption. [Note: unless, consistent with the RI-ISI methodology, a smaller break will result in more limiting consequences].
NRC Question #7 The Enclosure (Attachment 11) refers to the ASME RA-S-2002.Standard for Probabilistic Risk Assessment [PRA] for Nuclear Power Plant Applications, up through the RA-Sb-2005 Addenda, as an acceptable method for determining PRA scope, technical adequacy, and peer review requirements.
to 2CAN020804 Page 7 of 12 Is the ANO-2 PRA consistent with the standard up through the RA-Sb-2005 Addenda?
If the ANO-2 PRA has not been reviewed for consistency with the standard up through the RA-Sb-2005 Addenda, when was the last review performed and what guidelines were used for the review?
Please provide all the facts and observations identified during previous reviews of the ANO2 PRA and describe the resolution of these issues, or why the issues need not be resolved because resolution would have no or limited affect on the categorization.
Response
A Gap analysis to Regulatory Guide (RG) 1.200 requirements (including RA-Sb-2005 Addenda) was performed on the ANO-2 PRA in 2007. This was done to prepare the PRA for implementation into and development of Fire PRA models to support the transition to an NFPA 805-based fire protection licensing basis.
Except for three (3) B level gaps, all significant gaps were closed and the updated PRA model was provided to the Fire PRA development team.
The three B level gaps, which are in the process of being closed, relate to component boundary documentation, failure data, and documentation of walkdowns.
As discussed in the response to the NRCs question #8 below, all significant gaps identified by the earlier peer review of the ANO-2 PRA have been closed. Additionally, the remaining 3 model gaps, which are in the process of being closed, are not believed to be significant with respect to the RI-RRM application.
In addition to the above, as discussed in the response to question #9 below, the approved RI-ISI methodology has a commitment that Entergy will maintain and periodically update the program to reflect operating experience including updates to the PRA, as applicable.
As an example, this living component of a RI-ISI program, as well as a living RI-RRM program, requires that updates to the PRA be reflected in the supporting analysis (e.g. consequence assessment). This has been accomplished for the first inspection period for the RI-ISI application. As discussed in the response to NRC question #9 below, reflecting the latest PRA model at the time, had a negligible impact on the consequence ranking and therefore, the final results, for the RI-ISI application. Because the RI-RRM application is founded on the consequence assessment portion of the RI-ISI methodology, it is also expected that PRA updates will not have a significant impact on the RI-RRM program going forward.
From a timing perspective, the second RI-ISI update is about to begin and will also assess any impact of the most recent PRA information (e.g. the model revision discussed above) on the RI-ISI application as well as the RI-RRM application.
to 2CAN020804 Page 8 of 12 NRC Question #8 Please discuss the quality of the PRA that was used to support the proposed program and explain why this quality is sufficient to support the proposed changes to R&R activities.
Response
The ANO-2 Internal Events level 1 PSA Model has been through several model revisions since the approval of the RI-ISI program. Besides the update of initiating events and component failure data to reflect plant specific experience, the most important model element revisions are summarized below:
Initiating Event Fault Tree models Expanded the scope of the PSA (e.g., now includes ATWS, ISLOCA)
Human Reliability Analysis has been greatly improved to include dependencies between multiple HRA events in individual cutsets Common Cause Failure (CCF): Used the Entergy standard INEL CCF database &
methodology and calculated the uncertainty parameters Loss Off-Site Power (LOSP): Updated with the most recent EPRI data and used the EPRI convolution method to realistically credit recovery of offsite power System fault trees updated to address plant design and procedure changes Closed out numerous model change requests (MCRs) written to improve or correct errors in models In addition to the above, a Combustion Engineering Owners Group (CEOG) Peer Review of the ANO-2 model was performed in 2002. All significant Facts & Observations from this review were addressed during the model update(s).
A Gap Analysis to the ASME PSA Standard and RG 1.200 was performed on the ANO-2 PSA in 2007 (see the response to NRC question #7).
The responses to questions #7 and #9 also discuss the PRA used to support the original RI-ISI consequence assessment, status of the most recent ANO-2 PRA model as well as the impact of the most recent periodic RI-ISI update. As discussed in response to question #9, reflecting changes to the PRA into the consequence assessment has had a negligible impact on the RI-ISI consequence assessment and rankings. This is expected to be true for both the RI-ISI application and the RI-RRM application.
As discussed in the staffs Safety Evaluation on EPRI TR-112657 (i.e. the basis for the proposed RI-RRM methodology), the plant specific PRA is used to characterize specific attributes at the plant in a manner that can support and confirm the basic assumptions of the general methodology. Above and beyond PRA quality reviews (e.g. peer review, ASME RA-Sb-2005); the methodology itself provides means and criteria (e.g. look-up tables) that act to to 2CAN020804 Page 9 of 12 provide another level of PRA quality review. The methodology also includes systematic consideration of initiating events and operating states that may be outside the scope of the licensee's PRA such as external events and refueling operation. The methodology and application of the plant-specific PRA results are used to support placing pipe segments into broad consequence categories such that only large changes in the PRA would be expected to significantly impact the results in any meaningful manner.
Another important aspect of the EPRI consequence ranking approach that provides stability to the overall categorization process is that only PRA changes that result in an increase in a consequence rank would be of any significance. That is, PRA changes that change a consequence rank from Medium to Low would be conservative from a ranking perspective.
In addition to the above, section I-3.2.2(b) and (c) of the proposed methodology provides an additional level of review above and beyond the PRA and consequence assessments providing further assurance that implementation of the relief request results in a robust categorization.
To further support the robustness of the methodology and Entergys application of the methodology, much of the supporting analyses (e.g. consequence assessments) were submitted to the NRC for their review during the RI-ISI methodology development and approval process. NRCs approval of the methodology (EPRI TR-112657) and ANO-2s application of the methodology also concluded that large changes to the plant-specific PRA would typically be required to significantly impact the categorization results.
NRC Question #9 ANO-2 has implemented an update process for RI-ISI. Please identify any changes to the risk-informed update process that will be made in order to ensure that issues important to R&R activities will be included in the update.
Response
As noted above, as part of the ANO-2 RI-ISI approval, Entergy committed to periodically update the ANO-2 RI-ISI program in order to fulfill its RI-ISI living program commitment. Entergy has completed the first periodic review as required by this commitment. General observations from this review include:
IPE model formed the basis for the RI-ISI submittal the most recent PSA model at the time (PSA Model Rev. 3p2), formed the basis for the RI-ISI update which included changes to:
System updates, HRA updates Common cause updates Due to design changes, 8 welds were added to the scope of the RI-ISI program (all risk category 6, low risk) to 2CAN020804 Page 10 of 12 Several consequence rankings could change Existing plant programs are capturing operating experience (e.g. PWSCC)
Delta risk RI-ISI Submittal: risk decrease of 5E-08 (CDF)
RI-ISI Update: risk decrease of 1E-08 (CDF)
The specific conclusions drawn from this review, as well as their pertinence to the RI-RRM application, are presented below as defined by each step of the RI-ISI methodology:
Scope determination - the only impact identified was due to ER-ANO-2000-2804-011. The impact of this change was to add 8 risk category 6 welds to the scope of the RI-ISI program.
As these welds are risk category 6, no additional inspections are required. These will continue to be captured and reflected in support of the RI-RRM application.
Consequence evaluation - the updated PRA inputs were reviewed and no changes that would increase any consequence rankings were identified. These will continue to be captured and reflected in support of the RI-RRM application.
Degradation mechanism evaluation - applicable ERs, procedures, specifications, etc were reviewed and no changes to the RI-ISI evaluations were identified. These insights will continue to be captured and reflected in support of the RI-RRM application (e.g. are there any insights from these evaluations/experiences that would warrant different treatment practices).
Service history - NIS-2, FAC and MIC inspections, etc were reviewed and no changes to the RI-ISI evaluations were identified. These insights will continue to be captured and reflected in support of the RI-RRM application (e.g. are there any insights from these evaluations/experiences that would warrant different treatment practices).
Risk ranking - no changes to the base case risk ranking is required. Eight additional risk category 6 welds were added to the program scope as a result of ER-2000-2804-011. Not applicable to RI-RRM.
Element selection - no changes to the inspection population was required. Not applicable to RI-RRM.
Change-in-risk evaluation - the change-in-risk evaluation was updated to reflect the latest CCDP for LOCAs. The results of this update is that the RI-ISI program still produces a decrease in plant risk and meets NRC and EPRI TR-112657 acceptance criteria. This is not applicable to RI-RRM.
Monitoring - this review as well as existing plant practices (e.g. NIS-2, operating experience review) meet the intent of providing a monitoring mechanism for RI-ISI programs. These insights will continue to be captured and reflected in support of the RI-RRM application (e.g.
are there any insights from these evaluations/experience that would warrant different treatment practices).
to 2CAN020804 Page 11 of 12 NRC Question #10 The answers to questions 1, 2, 3 and 6 (on pages 7 and 8 of the Enclosure to the August 6, 2007, submittal (Reference 12)) did not provide adequate responses. Please address all parts of each question in your answers.
Response
The August 6, 2007, submittal (Reference 12) has been updated to respond to this RAI. This update is provided in Attachment 2.
REFERENCES
- 1. USNRC, NRC Leak-Before-Break Analysis Method for Circumferentially Through-Wall Cracked Pipes Under Axial Plus Bending Loads," NUREG/CR-4572, May 1986.
- 2. USNRC Generic Letter 90-05, Guidance for Performing Temporary non-Code Repair of ASME Code Class 1, 2, and 3 Piping.
- 3. USNRC Safety Evaluation Report, Arkansas Nuclear One, Units 1 and 2, Request for Generic Relief From ASME Section XI for Service Water System Microbiologically Induced Corrosion Flaws (TAC NOS. M97750, MA4919, MA2397 and MA3963), dated March 31, 1999.
- 4. USNRC Safety Evaluation Report, Joseph M Farley, Units 1 and 2, Request for Relief From ASME Code Repair Requirements For ASME Code Class 3 Piping (TAC NOS. MC2211 and MC2212), dated June 25, 2004.
- 5. USNRC NUREG/CR-4977, SHAG Test Series: Seismic Research on an Aged United States Gate Valve and on a Piping System in the Decommissioned Heissdampfreaktor (HDR).
- 7. EPRI, Generic Seismic Ruggedness of Power Plant Equipment, NP-5223.
- 8. EPRI, The October 1, 1987, Whittier Earthquake: Effects on Selected power, Industrial, and Commercial Facilities, NP-7126.
- 9. EPRI, The December 7, 1988, Armenia Earthquake: Effects on Selected Power, Industrial, and Commercial Facilities, NP-7359
- 10. Stevenson et al, Evaluation of the Reliability of Cold Piping Designed to ASME Boiler and Pressure Vessel Code Allowable as a Basis for Risk Informed Designs, SMIRT 16, Washington DC, August 2001.
- 11. Entergy letter to NRC dated April 17, 2007, Request for Alternative ANO2-R&R-004, Revision 1 Request to Use Risk-Informed Safety Classification and Treatment for Repair /
Replacement Activities in Class 2 and 3 Moderate Energy Systems (CNRO-2007-00015)
- 12. Entergy Letter to NRC dated August 6, 2007, Request for Alternative ANO2-R&R-004, Revision 1 Response to NRC Request for Additional Information (CNRO-2007-00028)
ATTACHMENT 1 COMPARISON OF ASME CODE CASE N-660, RISK-INFORMED SAFETY CLASSIFICATION FOR USE IN RISK-INFORMED REPAIR / REPLACEMENT ACTIVITIES AND THE PROPOSED METHODOLOGY to 2CAN020804 Page 1of 3 COMPARISON OF ASME CODE CASE N-660, RISK-INFORMED SAFETY CLASSIFICATION FOR USE IN RISK-INFORMED REPAIR / REPLACEMENT ACTIVITIES AND THE PROPOSED METHODOLOGY N660 Section N660 Proposed Methodology (1)
Inquiry Only defines classification criteria In addition to defining classification criteria also defines treatment requirements for LSS components (see 7th paragraph of section IV Basis for the Proposed Alternative, page 2 of 17, of the relief request) 1100 Scope Applies to Class 1, 2 and 3 High and Moderate Energy Systems This request is limited to Class 2 and 3 Moderate Energy Systems 1320 Required Disciplines While the proposed methodology does not explicitly call out these disciplines, they are necessary to implement the EPRI RI-ISI consequence assessment which is the foundation of the proposed methodology. As such, they have been and will continue to be fundamental to the relief request and its supporting analyses.
1330 Adequacy of the PRA This information is discussed in section IV Basis for the Proposed Alternative, page 2 of 17, of the relief request.
Section I-3.0.2 of the proposed methodology also discusses PRA technical adequacy, including the requirement to verify assumptions on equipment reliability for equipment not within the scope of the repair/replacement program (e.g. so called RISC-2 components). With respect to the ANO-2 application, it is also further discussed in response to RAI #7, #8 and #9.
to 2CAN020804 Page 2of 3 N660 Section N660 Proposed Methodology (1)
-9000 Glossary An updated glossary was provided in the November 15, 2006 submittal.
I-3.0 Consequence Assessment Expanded discussion provided in the proposed methodology.
I-3.1.1(g)
N/A Added section describing applicable configurations for when a failure should be postulated. This table is taken directly from EPRI TR-112657.
I-3.1.2(a)
Initiating Events (IE) Impact Group Assessment Added requirement that differences in the consequence rank between the use of Table I-1 and I-5 need to be reconciled.
I-3.1.2(b)
System Impact Group Assessment Added requirement that differences in the consequence rank between the use of Table I-2 and I-5 need to be reconciled.
Additionally, for defense in depth purposes added postulated failures that lead to zero defense (i.e. no backup trains) shall be assigned a High consequence.
I-3.1.2(c)
Combination Impact Group Assessment Added requirement that difference in the consequence rank between the use of Table I-3 and I-5 need to be reconciled.
I-3.1.2(d)
Containment Performance Simplified discussion.
I-3.1.2(e)
N/A Added text, taken from EPRI TR-112657, on shutdown events which was not included in N660.
I-3.1.2(f)
N/A Added text, taken from EPRI TR-112657, on external events which was not included in N660.
to 2CAN020804 Page 3of 3 N660 Section N660 Proposed Methodology (1)
I-3.1.3, I-3.1.4, I-3.1.5 Not modeled components, safety margin and defense in depth Reworded and moved to I-3.2.2 I-3.2.2(b)
Any piping segment initially determined to be a Medium consequence category and that is subject to a known active degradation mechanism shall be classified as HSS.
Requirement deleted for the following reasons:
The consequence assessment evaluates a spectrum of break sizes from small to large. This is done with an assumed failure probability of 1.0. As such, this assessment envelopes any impact that a postulated degradation mechanism could have.
Operating conditions (e.g. temperature, pressure) coupled with a well engineered design (i.e. national standards) assures an incredibly low likelihood of large pressure boundary failures for moderate energy systems. This has been demonstrated, even with degraded conditions (e.g. wall thinning, through wall flaws), in numerous industry and USNRC documents
[References 1 - 10]. These documents and their technical bases provide a high degree of confidence that large pressure boundary failure in moderate energy systems is extremely unlikely.
Finally, the treatment requirements of this relief assure that LSS components will continue to reliably perform their safety related functions under design basis conditions.
(1) CNRO-2007-00015, dated April 17, 2007,
ATTACHMENT 2 REVISED RESPONSE TO QUESTIONS 1, 2, 3, AND 6 OF THE AUGUST 6, 2007 SUBMITTAL to 2CAN020804 Page 1 of 2 NRC Question #1 Failure of the pressure retaining function of the segment will not directly or indirectly (e.g.,
through spatial effects) fail a basic safety function?
Response
Per the proposed methodology, which is founded on section 3.3 of EPRI TR-112657, it is a requirement to assess all safety functions supported by the subject system. The consequence ranking for each individual segment is then a function of the most limiting situation (e.g. injection versus recirculation). Per Table 1 above, any segment classified as Medium or Low will have at least one train unaffected by the postulated break for all functions supported by the system. Per the glossary, loss of a single train would typically not constitute loss of a basic function. Therefore, failure of any segment classified as Medium or Low will not fail a basic safety function.
NRC Question #2 Failure of the pressure retaining function of the segment will not prevent the plant from reaching or maintaining safe shutdown conditions; and the pressure retaining function is not significant to safety during mode changes or shutdown. Assume that the plant would be unable to reach or maintain safe shutdown conditions if a pressure boundary failure results in the need for actions outside of plant procedures or available backup plant mitigative features.
Response
Although the CSS can be used to support shutdown (e.g. as an alternative to LPSI pumps, RWT provides inventory for filling refueling canal), the CSS is typically not used to support mode changes or shutdown cooling. Per the proposed methodology, the impact of CSS segment failures on plant shutdown events was evaluated. The results of the consequence evaluations conducted, identified that any segment classified as Medium or Low will have at least one train unaffected by the postulated break. The consequence evaluation did not credit operator action outside of those that are proceduralized.
NRC Question #3 The pressure retaining function of the segment is not called out or relied upon in the plant Emergency/Abnormal Operating Procedures or similar guidance as the sole means for the successful performance of operator actions required to mitigate an accident or transient.
Response
The consequence assessment requires that both direct and indirect effects be considered on the system under evaluation (i.e. the CSS) as well as any other potentially effected systems /
components (e.g. due to spray, loss of inventory). While CSS is credited in the Emergency Operating Procedures (EOPs), per the consequence evaluations conducted above, any segment classified as Medium or Low will have at least one train unaffected by the postulated to 2CAN020804 Page 2 of 2 break. As such, these segments (i.e. those categorized as Medium or Low) are not the sole means for successful performance to mitigate an accident or transient.
NRC Question #6 Reasonable balance is preserved among prevention of core damage, prevention of containment failure or bypass, and mitigation of an offsite release.
Response
This balance is preserved as there is no change to the design, design basis or operation of the CSS by this change. Additionally, the consequence assessment of the proposed methodology (see section 3.3. of EPRI TR-112657) requires an evaluation and ranking of postulated CSS failures on core damage and containment performance (e.g. bypass, LERF).
Finally, with implementation of this relief request, the CSS will still be required to reliably perform its safety-related function.
ENCLOSURE 2 2CAN020804 LIST OF REGULATORY COMMITMENTS to 2CAN020804 Page 1 of 1 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Entergy in this document.
Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
TYPE (Check one)
COMMITMENT ONE-TIME ACTION CONTINUING COMPLIANCE SCHEDULED COMPLETION DATE Entergy shall review changes to the plant, operational practices, applicable plant and industry operational experience, and, as appropriate, update the probabilistic risk assessment (PRA) and categorization and treatment processes. Entergy shall perform this review in a timely manner but no longer than once every two refueling outages.
X Upon implementation of ANO2-R&R-004, Revision 1 The analyses performed in support of R&R activities will follow the requirement in the proposed methodology for crediting operator action.
X Upon implementation of ANO2-R&R-004, Revision 1