ML090930246

Approval of Request for Alternative ANO2-R&R-004, Revision 1, to Use Risk-Informed Safety Classification and Treatment Repair/Replacement Activities in Class 2 & 3 Moderate Energy Systems

ML090930246
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/22/2009
From:	Markley M Plant Licensing Branch IV
To:	Entergy Operations
Wang, A B, NRR/DORL/LPLIV, 415-1445
References
ANO2-R&R-004, Rev 1, TAC MD5250
Download: ML090930246 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 22, 2009 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT 2 - APPROVAL OF REQUEST FOR ALTERNATIVE AN02-R&R-004, REVISION 1, REQUEST TO USE RISK-INFORMED SAFETY CLASSIFICATION AND TREATMENT FOR REPAIR/REPLACEMENT ACTIVITIES IN CLASS 2 AND 3 MODERATE AND HIGH ENERGY SYSTEMS (TAC NO. MD5250)

Dear Sir or Madam:

By letter dated April 17, 2007, as supplemented by letters dated August 6, 2007, February 20, 2008, and January 12, 2009, Entergy Operations Inc. (Entergy, the licensee), submitted a request for alternative, AN02-R&R-004, Revision 1, to the U.S. Nuclear RegUlatory Commission (NRC) staff. Entergy requested that the NRC authorize the use of a risk-informed safety classification and treatment program for repair/replacement activities in Class 2 and 3 systems at Arkansas Nuclear One, Unit 2 (ANO-2). The request is applicable to the remainder of the third and fourth 1O-year inservice inspection (lSI) intervals.

The NRC staff completed its review of the subject request for alternative. Based on the enclosed safety evaluation (SE), the NRC staff has determined that the proposed alternative, AN02-R&R-004, Revision 1, to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, IWA-4000 to use a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items or their associated supports (exclusive of Class CC and MC items), provides an acceptable level of quality and safety. Therefore, pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations, the NRC staff authorizes the use of the alternative AN02-R&R-004, Revision 1 for ANO-2 for the remainder of the third and the fourth 10-year lSI interval.

All other requirements of the ASME Code,Section XI, for which relief has not been specifically requested remain applicable, including a third-party review by the Authorized Nuclear Inservice Inspector.

- 2 If you have any questions regarding the SE, please contact Alan B. Wang at (301) 415-1445.

Sincerely, Michael 1. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-368

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE AN02-R&R-004, REVISION 1, REQUEST TO USE RISK-INFORMED SAFETY CLASSIFICATION AND TREATMENT FOR REPAIR/REPLACEMENT ACTIVITIES IN CLASS 2 AND 3 MODERATE AND HIGH ENERGY SYSTEMS THIRD AND FOURTH 10-YEAR INSERVICE INSPECTION INTERVALS ENTERGY OPERATIONS, INC ARKANSAS NUCLEAR ONE, UNIT 2 DOCKET NO. 50-368

1.0 INTRODUCTION AND BACKGROUND

By letter dated April 17, 2007 (Reference 1), as supplemented by letters dated August 6, 2007 (Reference 2), February 20, 2008 (Reference 3), and January 12, 2009 (Reference 4)

(Agencywide Documents Access and Management System (ADAMS) Accession Nos.

ML071150108, ML072220160, ML080520186, and IVIL090120620, respectively), Entergy Operations Inc. (Entergy), submitted request for alternative, AN02-R&R-004, Revision 1, for U.S. Nuclear Regulatory Commission (NRC) staff review and approval. In Reference 4, Entergy modified its request to include both moderate and high energy Class 2 and 3 systems.

In its submittal, Entergy proposed an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, IWA-4000 to use a risk-informed safety classification and treatment program for repair/replacement activities in Class 2 and 3 systems at Arkansas Nuclear One, Unit 2 (ANO-2). The risk-informed safety classification process would separate structures, systems, and components (SSCs) into high safety-significant (HSS) and low safety-significant (LSS) populations. Repair/replacement activities that will be applied to the LSS SSCs may be modified from those activities otherwise required by the regulations. Reference 1 also provides a discussion of the repair/replacement treatment that would be applied to the LSS SSCs as an alternative to the current requirements.

The submittal describes the risk-informed safety categorization process that Entergy proposes to use at ANO-2 to determine the risk-informed safety class (RISC) for Class 2 and 3 pressure retaining items or their associated supports (exclusive of Class CC and MC items). In Reference 1, Entergy proposed to limit the categorization to moderate energy systems.

However, Reference 4 proposed to expand the categorization to high energy systems. Entergy clarified that the proposed categorization method consists, in essence, of the consequence Enclosure

- 2 assessment portion of its NRC staff approved (Reference 5) risk-informed inservice inspection (RI-ISI) methodology, supplemented with the "additional considerations" contained in Nuclear Energy Institute (NEI) 00-04, "10 CFR [Title 10 of the Code of Federal Regulations] 50.69 SSC Categorization Guideline" (Reference 6). Entergy provided an example application of the risk-informed safety classification (categorization) process for the containment spray system (CSS), in its enclosure to Reference 2. Upon approval of the relief request, Entergy will conduct the additional evaluations required to supplement the RI-ISI categorization process to establish the consequence ranking assignments for Class 2 and 3 SSCs before using risk-informed repair/replacement activities.

Entergy's proposed categorization methodology is based, in large part, on Entergy's RI-ISI methodology. The NRC does not endorse Entergy's methodology for generic use. In its letter dated January 12, 2009, submittal (Reference 4), Entergy refers to its proposal as a risk-informed repair/replacement application (RI-RRA). This safety evaluation refers to the description of the final method (as described in Attachment 1 of Reference 1 as further modified in References 2,3, and 4) as the RI-RRA method.

2.0 REGULATORY EVALUATION

Paragraph 50.55a(g) of Title 10 of the Code of Federal Regulations (10 CFR) specifies that inservice inspection (lSI) of nuclear power plant components shall be performed in accordance with the requirements of the ASME Code except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Paragraph 50.55a(a)(3) of 10 CFR states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASIVIE Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of components. The applicable Code of record for repair/replacements for the third 10-year lSI interval for ANO-2, is the ASME Code,Section XI, 2001 Edition through the 2003 Addenda.

On November 22, 2004, the Commission adopted a new Section 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," of 10 CFR on risk-informed categorization and treatment of SSCs for nuclear power plants (69 FR 68047). This new section permits power reactor licensees and license applicants to implement an alternative regulatory framework with respect to "special treatment" of LSS SSCs. Special treatment refers to those requirements that provide increased assurance (beyond normal industrial practices) that SSCs perform their design-basis functions. In May 2006, the NRC staff issued Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components In Nuclear Power Plants According To Their Safety Significance, For Trial Use," Revision 1 (Reference 7). RG 1.201 endorses a categorization method, with conditions, for categorizing active SSCs described in NEI 00-04. Entergy has not

- 3 requested to implement 50.69, but instead has requested to categorize passive SSCs (e.g.,

piping) and implement alternative special treatment activities limited to the repair/replacement activities for Class 2 and 3 pressure retaining items or their associated supports (exclusive of Class CC and MC items).

In July 2002, the ASME issued Code Case N-660, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1," (Reference 8). Code Case N-660 (N-660) described how the safety significance of pressure retaining equipment (e.g., piping) could be categorized in order to change the repair/replacement requirements for the LSS piping. In October 2007, the NRC staff issued RG 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," (Reference 9). In RG 1.147, Revision 15, the NRC conditionally accepted N-660. The NRC staff condition is that, "The Code Case must be applied only to ASME Code Classes 2 and 3, and non-Code Class pressure retaining components and their associated supports." By endorsing N-660, the NRC staff endorsed a categorization methodology that may be used to determine the risk-informed safety classification for passive SSCs. Once the categories have been developed according to the endorsed method, alternative repair/replacement activities may be used for LSS SSCs instead of the ASME requirements. Since N-660 was developed, trial applications of this Code case have been conducted. Lessons learned from these trial applications have resulted in various attempts to revise N-660. N-660 is generally considered to be very conservative, in part, due to the fact that the process places essentially all passive equipment into the HSS category.

In November 2002, the NRC issued RG 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 10). RG 1.174 provides that probabilistic risk assessment (PRA) capability should be commensurate with the regulatory application, and provides guidance for demonstration of sufficient PRA capability. N-660 states that the licensee is responsible for demonstrating the adequacy of the PRA used as a basis for the classification. On March 22, 2007, the NRC issued Regulatory Issue Summary (RIS) 2007-06, "Regulatory Guide 1.200 Implementation."

The RIS informed licensees that the NRC staff will use Revision 1 of RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Reference 11), to assess technical adequacy for all risk-informed applications received after December 2007. RG 1.200 addresses the use of the ASME RA-Sa-2005, Addenda to ASI\\IIE RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 12), and the NEI peer review process NEI 00-02, "PRA Peer Review Process Guidance" (Reference 13). This application was submitted by letter dated April 17, 2007, and was, therefore, submitted before RG 1.200 was scheduled to be applied. The quality of the PRA used to support this submittal is, however, evaluated against the guidance in RG 1.200.

Entergy proposes a safety classification methodology that differs from the methodology described in N-660. The NRC staff evaluated the acceptability of this alternative method based, in part, on consistency of the proposed changes to N-660 with the requirements in 10 CFR 50.69, with categorization guidelines in NEI 00-04 as endorsed in RG 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," and with the generic risk-informed decision-making guidelines established in RG 1.174.

- 4 2.1 Monitoring of HSS and LSS SSCs 10 CFR 50.69(e)(1) requires that licensees review changes to the plant, operational practices, applicable plant and industry operational experience and, as appropriate, update the PRA and SSC categorization and treatment processes. Safety-significant categories RISC-1 and RISC-2 in 10 CFR 50.69 are equivalent to the licensee's HSS category. Safety-significant categories RISC-3 and RISC-4 in 10 CFR 50.69 are equivalent to the licensee's LSS category. In 10 CFR 50.69(e)(2), licensees are required to monitor the performance of RISC-1 and RISC-2 SSCs and make adjustments as necessary to either the categorization or treatment processes so that the categorization process and results are maintained valid. In addition, all safety-related SSCs are classified RISC-1 or RISC-3 and, therefore, are subject to the lSI and inservice testing (1ST) requirements in 10 CFR 50.55a, "Codes and Standards," and the quality assurance requirements in 10 CFR Part 50, Appendix S, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," including Criterion XVI, "Corrective Action."

Under 10 CFR 50.69(d)(2)(i), licensees are required to conduct periodic inspections and tests to determine that RISC-3 SSCs will remain capable of performing their safety-related functions under design-basis conditions. In addition, 10 CFR 50.69(d)(2)(ii) requires that conditions that would prevent a RISC-3 SSC from performing its safety-related functions under design-basis conditions be corrected in a timely manner and, that for significant conditions adverse to quality, measures be taken to provide reasonable confidence that the cause of the condition is determined and corrective action taken to preclude repetition.

Specifically for RISC-3 SSCs, 10 CFR 50.69(e)(3) requires that licensees consider data collected in 10 CFR 50.69(d)(2)(i) to determine ifthere are any adverse changes in performance such that the SSC unreliability values approach or exceed the values used in the evaluations to satisfy 10 CFR 50.69(c)(1 )(iv). The licensee shall make adjustments as necessary to the categorization or treatment processes so that the categorization and results are maintained valid.

Furthermore, 10 CFR 50.69(c)(1 )(iv) requires that, for RISC-3 SSCs, the categorization process must include evaluations that provide reasonable confidence that sufficient safety margins are maintained and that any potential increases in core damage frequency (CDF) and large early release frequency (LERF) resulting from changes in treatment are small.

2.2 Application of HSS and LSS Treatment Requirements As stated above, 10 CFR 50.69 permits power reactor licensees and license applicants to implement an alternative regulatory framework for the "special treatment" of LSS SSCs. Special treatment refers to those requirements that provide increased assurance beyond normal industrial practices that SSCs will perform their design-basis functions.

The regulation at 10 CFR 50.69(d)(2) requires that licensees or applicants ensure, with reasonable confidence, that RISC-3 SSCs remain capable of performing their safety-related functions under design-basis conditions, including seismic conditions and environmental conditions and effects throughout their service life. The treatment of RISC-3 SSCs must be

- 5 consistent with the categorization process. Inspection and testing, and corrective action shall be provided for RISC-3 SSCs.

In August 2002, the ASME issued Code Case N-662, "Alternative Repair/Replacement Requirements for Items Classified in Accordance With Risk-Informed Processes,Section IX, Division 1" (Reference 14). The Code case described alternative repair/replacement requirements for items classified in accordance with risk-informed classification criteria. In Revision 15 of RG 1.147 (Reference 9), the NRC conditionally accepted Code Case N-662.

The NRC staff condition for the use of N-662 is that, "The Code Case must be applied only to ASME Code Classes 2 and 3, and non-Code Class pressure retaining components and their associated supports."

3.0 TECHNICAL EVALUATION

Entergy proposed its original RI-RRA methodology in Attachment 1, (Reference 1) which would replace, in its entirety, N-660. In its enclosure to Reference 2, Entergy also provided an example application of the risk-informed classification process for its CSS. The example application expands and clarifies many of the guidelines in Attachment 1. In Reference 4, Entergy provided a table summarizing the proposed changes between N-660 and its final proposed RI-RRA methodology. The RI-RRA methodology is based on, and very similar to, the methodology described in N-660.

For example, all the tables from N-660 used for the categorization of SSC are also included in the RI-RRA method. Many of the proposed changes are editorial changes or reorganization changes to the content. Entergy reorganized the sections dealing with the treatment of various qualitative considerations which resulted in numerous individual editorial changes. The NRC staff concludes that all of the editorial changes do not affect the safety-significant classification assigned to the SSCs, and are, therefore, acceptable.

The licensee proposed several substantive changes to the methodology in N-660 that will affect the safety-significant classification assigned to SSCs. Proposed substantive changes include clarifying how operator actions are credited; clarifying how shutdown operation and external initiating events are characterized; modifications to the qualitative considerations; and guidelines for basing the consequence evolution on small instead of large breaks. The licensee also described its plant-specific implementation of N-660's generic PRA quality and monitoring requirements and described its proposed treatment program. Each of these issues is discussed below.

3.1 Credit for Operator Actions Code Case N-660 permits consideration of possible operator actions in Sections 1-3.1.1 (b),

1-3.1.1 (e), and 1-3.1.3(a)(3) but provides no clarification which operator actions may be credited or how they should be credited. The RI-RRA method Section 1-3.0.1 provides additional guidance about crediting operator action by stating that:

Throughout the evaluations specified in Sections 1-3.0, 1-3.1, and 1-3.2, credit may be taken for plant features and operator actions to the extent these would not be affected by

- 6 failure of the segment under consideration. When crediting operator action, the likelihood for success and failure will be determined consistent with ANO-2's NRC-approved RI-ISI application. The scenario that results in the highest consequence ranking shall be used.

As part of its September 30, 1997, RI-ISI submittal (Reference 15), the licensee describes in detail how operator actions are credited in the RI-ISI evaluation (Reference 16). Operator actions introduced during the RI-ISI evaluation are generally limited to actions to isolate a failed segment by closing motor or air operated valves using controls located in the control room. 1 Isolating a failed segment reduces the number of SSCs that fail because of direct or indirect effects and can change the accident scenario. For example, successful isolation could change a scenario from a main feedwater (MFW) plant trip with consequential loss of both auxiliary feedwater water (AFW) trains to a MFW plant trip with loss of one AFW train. The likelihood of the failure of each postulated operator action is developed using scenario-specific influences on human performance. Two risk estimates are then produced for each segment failure that could be isolated. The first estimate develops the risk from the accident scenario where the segment remains unisolated, multiplied by the likelihood that the operator fails to isolate. The second estimate develops the risk from the accident scenario where the segment is successfully isolated. The higher of the two risk estimates is assigned to the segment and used in the safety-significant classification.

The likelihood of the failure of each postulated operator action is developed using scenario specific influences on human performance, consistent with generally acceptable PRA practice.

Evaluating the scenarios that include both successful and unsuccessful operator actions ensures that all scenarios are evaluated despite the methodology not requiring quantification of event tree structures that would otherwise provide this assurance. The NRC staff concludes that the licensee's incorporation of operator actions into the classification process is acceptable because the methodology should identify HSS segments with appropriate consideration of possible operator actions.

3.2 Evaluation of Shutdown Operation and External Initiating Events Section 1-3.0 in N-660 directed the licensee to ensure that the consequence assessment includes information for "each piping segment that is not modeled in the PRA, but considered relevant to the classification (e.g., information regarding design-basis accidents, shutdown risk, containment isolation, flooding, fires, seismic conditions)." The RI-RRA method adds Sections 1-3.1.2(e) and (f) that discuss in greater detail what should be addressed while evaluating shutdown operations and external initiating events, respectively. The guidance provided in (e) directs the licensee to identify the different safety functions and success criteria in the different stages of all modes of operation, and to evaluate the effects of pressure boundary failure during non-power operation. The guidance in (f) directs the licensee to evaluate the effect of external events that can cause a pressure boundary failure or create demands that might cause a pressure boundary failure, and to determine whether any LSS segments should be assigned as HSS based on this evaluation.

1 In Reference 16, the licensee stated that there are a few cases where isolation could be performed locally but that credit was not taken for these actions.

- 7 The additional guidance provided in the RI-RRA method is an improvement over N-660 which simply directed the licensee to address these issues. The NRC staff finds that identifying the changing functional requirements between power and non-power operation and then considering the impact of segment failures on these non-power functional requirements should identify LSS segments that become HSS during shutdown operation. The NRC staff also finds that the external event evaluation should identify LSS segments that should be HSS based on their sensitivity to, or contribution to mitigating, external initiating events. Therefore, the NRC staff concludes that RI-RRA method is an acceptable method to determine the impact of shutdown operation and external events that are not modeled in the PRA.

3.3 Modification of Qualitative Considerations N-660 describes a two stage classification process. In the first stage, PRA analyses (or a series of tables which are equivalent to PRA analyses) are used to identify HSS SSCs. In the second stage, licensee personnel reevaluate the remaining potentially LSS segments. For each segment, qualitative considerations are addressed through a series of conditions or questions which can be assigned a "true" or "false" status for each piping segment. The response to these questions support the systematic determination on whether SSCs that are not assigned HSS by the quantitative PRA results should, nevertheless, be assigned HSS based on qualitative considerations. A similar process, and similar questions, is included in the NEI 00-04 methodology, endorsed in RG 1.201, for safety significance classification of active SSCs to support the implementation of 10 CFR 50.69. As discussed below, the licensee's RI-RRA methodology rearranged some N-660 questions, deleted some questions, and modified some questions.

3.3.1 Rearranged Questions All the questions were reformulated in the RI-RRA method to elicit a "false" response instead of a "true" response if the undesired characteristic was present. The criterion for assigning the segment as HSS was correspondingly modified such that any "false" response should lead to the segment being assigned HSS. This change brings the process into better alignment with the general classification process described in the I\\JEI 00-04 and is acceptable because there are no technical changes.

3.3.2 Deleted Questions The question in N-660 Section 1-3.1.3(a)(1) was deleted. The response to this question would designate a segment as HSS if its failure would significantly increase the frequency of an initiating event. The RI-RRA method categorizes the passive functions of SSCs assuming that the segment has failed and determining the safety significance of the consequence of the failure. All the effects of a pipe rupture, including all initiating events it causes, are already addressed as part of the categorization process. Therefore, the NRC staff concludes that the question is redundant and, therefore, the proposed deletion is acceptable.

The question in N-660 Section 1-3.1.3(a)(2) was deleted. The response to this question would require that all Class 1 items (except for some Class 1 parts defined in the 50.55a(c)(2)(i) and

- 8 (ii)) should be classified as HSS. Since Entergy's methodology is only applied to Class 2 and Class 3 piping the NRC staff has concluded that this question is not applicable and, therefore, the proposed deletion is acceptable.

The question in N-660 Section 1-3.1.3(b)(1) was deleted. The response to this question would require that all piping in every system that supports the retention of fission products during severe accidents be assigned to the HSS category. The NRC staff agrees that the N-660 guidance would place whole systems into the HSS category based on small, and perhaps very small, parts of the system acting as a barrier to fission product release. All of the effects of piping rupture, including the potential to cause or permit a release during a severe accident, are addressed as part of the passive categorization process. The conditional large early release probability (CLERP) guidelines should identify all HSS piping segments in a system whose failure contributes significantly to fission product release. The NRC staff concludes that the question in N-660 Section 1-3.1.3(b)(1) is excessively conservative and, therefore, the proposed deletion is acceptable.

The direction in N-660 Section 1-3.2.2(b) was deleted. The section stated that any piping segment determined to be potentially an LSS segment but that is subject to a known active degradation mechanism shall be classified as HSS. Entergy will replace the existing Section XI repair/replacement requirements with owner-defined activities to confirm with reasonable confidence that each LSS item remains capable of performing its safety function(s) under design-basis conditions. Any condition identified that would prevent an LSS component from performing its safety-related function(s) under design-basis conditions will be corrected in a timely manner. Therefore, any known active degradation mechanism in any Class 2 or Class 3 piping classified as LSS shall be evaluated by the licensee and, if its presence causes a significant condition adverse to quality, it would be corrected via Entergy's corrective action program. Therefore, the NRC staff concludes that "known active degradation" is appropriately addressed in treatment and need not be considered during classification.

3.3.3 Modified Questions The questions in N-660 Sections 1-3.1.3(a)(3) and 1-3.1.3(a)(4) were combined into the RI-RRA method Section 1-3.2.2(b)(1). N-660 separately questioned whether consequential failure of other safety-significant functions or other safety-significant segments is expected. The proposed method combines these into a single question about whether a consequential failure of a basic safety function is expected. This change is acceptable because accident scenarios are defined by failures of functions (which may be caused by equipment failure) and because "basic safety functions" are expected to include all safety-significant functions as well as some functions that might not be safety significant.

The question in N-660 Section 1-3.1.3(a)(5) was modified by removing "adversely affect" from "prevent or adversely affect" the plant from reaching or maintaining safe shutdown. In response to the question in Section 1-3.2.2(b) of the RI-RRA method, individual segment failures which could prevent the plant from reaching safe-shutdown will continue to be assigned HSS. Piping failures which only "adversely affect" a plant's capability to reach and maintain safe-shutdown are appropriately addressed as the possible loss of a basic safety function in Section 1-3.2.2(b)(1) of the RI-RRA method and need not be reconsidered in this question.

- 9 The question in N-660 Section 1-3.1.3(b)(2) was modified to address the operability of instrumentation that is "solely" relied upon by the operators instead of generally "support" operator actions to mitigate a transient (e.g., prevent core damage). Additionally, a new question was added in the RI-RRA Section 1-3.2.2(b)(4) which addresses the possible impact of segment failure on long-term containment integrity. Finally, the question in N-660 Section 1-3.1.3(b)(3) was modified to refer to radioactive material releases that cause implementation of offsite emergency planning activities instead of releases exceeding 10 CFR Part 100, Reactor Site Criteria," limits. All three changes emphasize identifying piping failures that cause safety significant consequences are consistent with the guidance provided in NEI 00-04 and, therefore, are acceptable.

The licensee also expanded N-660 Sections 1-3.1.4 and 1-3.1.5 which discuss maintaining defense-in-depth philosophy and sufficient safety margins, respectively. The expanded discussions direct the licensee personnel to verify that assigning each segment to the LSS category is not contrary to characteristics of maintaining defense-in-depth philosophy and sufficient safety margins identified in Sections 1-3.2.2(b)(6) through (10) and 1-3.2.2(c) in the Rl-RRA method. The characteristics listed in the RI-RRA method are consistent with those discussed in RG 1.174 and in NEI 00-04 and, therefore, are acceptable.

3.4 Consequence Evaluation Based on Small Break Size Section 1-3-1.1 (a) in N-660 required that the consequence analysis be performed assuming a large pressure boundary failure unless one or more of the three criteria listed in the Code case could be met. If anyone of these criteria was met, a smaller break could be assumed when determining the affects of the pressure boundary failure. Smaller breaks tend to result in damage to fewer nearby SSCs and slower transients than larger breaks. Assessing the consequence for small instead of large breaks could result in assigning a lower safety significance to pressure boundary failures.

The first of the three criteria in N-660 simply permits the consequences of a smaller leak to be used if more conservative than using a larger break. The second and third criteria, when met, provide confidence that a large break is very unlikely according to NRC-endorsed methods regardless of how the piping in question is repaired or replaced.

In Reference 1, the licensee proposed to modify the criteria to allow smaller pipe breaks to be assumed while developing the consequences, and to only apply the methodology to piping in moderate energy systems. In response to NRC questions, the licensee proposed changing the categorization process to consider large pipe breaks as described in N-660, and to increase the scope of categorization from only moderate energy systems, to both moderate and high energy systems. The final Rl-RRA method, described in Reference 4, confirms with the N-660 criteria describing the conditions when a smaller pipe break can be assumed and also expands the scope of piping which may be categorized to all Class 2 and 3 piping (excluding Class CC and MC items) as endorsed in RG 1.147. The NRC staff concludes that the final RI-RRA method is consistent with the method accepted by the NRC staff in N-660 and, therefore, is acceptable.

- 10 3.5 PRA Quality N-660 states that the licensee is responsible for demonstrating adequacy of any PRA used as the basis for this process. All deficiencies identified shall be reconciled during the analysis to support the RISC process and the resolution of all PRA issues shall be documented. Section 1-3.0.2 in the RI-RRA methodology states that:

The technical adequacy of the PRA used to support the evaluations required by this attachment shall be assessed. The PRA technical adequacy basis for the ANO-2 application shall be reviewed to confirm it is applicable to the safety significant categorization of this application, including verifying assumptions on equipment reliability for equipment not within the scope of this request.

In Reference 3, the licensee describes its assessment of the quality of its internal events PRA model. A Combustion Engineering Owners Group Peer Review of the ANO-2 model was performed in 2002. All significant facts and observations from this review were addressed during the model update(s). Appendix B in RG 1.200 endorses the NEI peer review process described in NEI-00-02 with clarifications and qualifications (i.e., a "gap" analysis). The licensee reported that it completed a gap analysis to RG 1.200 requirements (including RA-Sb-2005 Addenda) in 2007. Except for three B-level gaps, all significant gaps were closed. The three B level gaps, which are in the process of being closed, relate to component boundary documentation, failure data, and documentation of walkdowns. The licensee stated and the NRC staff agrees that these remaining three model gaps, which are in the process of being closed, are not believed to be significant with respect to the application.

In References 2 and 3, the licensee summarized the results of its evaluation of non-power operations and external events. The licensee concluded that the consequence ranking of the segments during shutdown operation would be no more risk significant than the ranking identified during power operation. The licensee also described the results of its review of the potential importance of failures during seismic and fire events. The licensee evaluated the consequences of these events by evaluating the number of available trains, frequency of challenge, and exposure time. This methodology forms the basis for Table 1-2 through 1-4 in N-660 and is, therefore, acceptable. The licensee identified no LSS segments should become HSS because of the risk from external events or non-power operation.

The RI-ISI consequence assessment for all 2 Class 2 and Class 3 piping at ANO-2 was completed in support of its RI-ISI relief request dated September 30, 1997. In its letter dated December 29, 1998 (Reference 5), the NRC staff concluded that the consequence methodology as applied by the licensee at ANO-2, is consistent with the conditional core damage probability (CCDP) and CLERP guidelines (i.e., is capable of classifying the segments according to the guideline values). The licensee reported in Reference 3 that it completed its periodic update of its RI-ISI program using its PRA model "Revision 3p2." As part of this update, the updated PRA 2 In some cases, Class 2 and Class 3 piping was screened out prior to the RI-ISI categorization. In Reference 3, the licensee stated that any system or portion of system that used the screening process or was not evaluated in detail will be classified as HSS unless subsequently evaluated with the RI-RRA methodology.

- 11 inputs were reviewed and no changes that would increase any consequence ranking were identified.

NRC RG 1.174 provides that PRA capability should be commensurate with the regulatory application. As discussed in the NRC staff's Safety Evaluation contained in Electric Power Research Institute (EPRI) Topical Report TR-112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, Final Report" (Reference 17), which describes the RI-ISI consequence evaluation that is the basis for the methodology, the plant-specific PRA is primarily used to characterize plant-specific attributes such as mitigating train reliability and human failure probabilities in a manner that can support and confirm assigning piping system failures consequences into broad safety-significant categories. Above and beyond PRA quality reviews, the methodology itself provides means and criteria (e.g. look-up tables) that act to 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.

The NRC staff concludes that the licensee's PRA analysis, supplemented with the external events and shutdown evaluations, will appropriately identify segments of high safety significance (i.e., segments with a CCDP and CLERP greater than 10-4 and 10-5, respectively, or segments which qualitative considerations indicate should be HSS) and is, therefore, acceptable.

3.6 Monitoring of HSS and LSS SSCs In Reference 1, Entergy states that they will review changes to the plant, operational practices and industry operational experience and, as appropriate, update the PRA and the categorization and treatment processes. This review shall be performed in a timely manner but no longer than once every two refueling outages. Reference 1 also states that any condition identified that would prevent an LSS component from performing its safety-related function(s) under design basis conditions will be corrected in a timely manner. For significant conditions adverse to quality that may be identified, measures will be taken via Entergy's corrective action program, established in accordance with 10 CFR Part 50, Appendix B, to provide reasonable assurance that the cause of the condition is determined and corrective action taken to preclude repetition.

RISC-1 SSCs (safety-related HSS) and RISC-3 (safety-related LSS) are subject to the regulatory requirements for safety-related equipment specified in 10 CFR Part 50. For example, SSCs within the scope of the ASIVIE Code and the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) are required to meet the lSI and 1ST requirements specified in 10 CFR 50.55a. Among those requirements is the 1ST provision for periodically assessing the operational readiness of pumps and valves to perform their safety functions, and the lSI provisions that require a mandatory program of examinations, pressure testing, and inspections for determining component acceptability for continued service and to manage deterioration and aging effects, along with repair/replacement activity requirements. Further, Quality Assurance Criterion XVI, "Corrective Action," in 10 CFR Part 50, Appendix B, states that measures shall be

- 12 established to assure that conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, Criterion XVI requires that the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The licensee's proposed alternative has not requested any changes to these requirements.

The NRC staff concludes that monitoring of HSS and LSS SSCs, as specified with the safety-related requirements for monitoring RISC-1 and RISC-3 SSCs, provides an adequate means of monitoring these SSCs such that the results of this monitoring can be used to adjust the categorization or treatment processes so that the categorization and results are maintained valid. Therefore, the NRC staff concludes that the monitoring, in accordance with safety-related requirements for RISC-1 (safety-related HSS) and RISC-3 (safety-related LSS) SSCs is consistent with the monitoring requirements of 10 CFR 50.69.

3.7 Application of HSS and LSS Treatment Requirements Entergy proposes alternative repair/replacement requirements for items classified in accordance with risk-informed classification that differs from the requirements described in Code Case N-662. The NRC staff evaluated the acceptability of these alternative requirements based on consistency with the requirements in 10 CFR 50.69.

Entergy's proposed alternative states that components categorized as HSS will continue to meet the ASME Code,Section XI requirements for repair/replacement activities. Those components categorized as LSS will be exempt from the ASME Code requirements for repair/replacement activities. Entergy will replace the existing Section XI repair/replacement requirements with owner-defined activities to confirm with reasonable confidence that each LSS item remains capable of performing its safety function(s) under design-basis conditions. Any condition identified that would prevent an LSS component from performing its safety-related function(s) under design-basis conditions will be corrected in a timely manner. For significant conditions adverse to quality that may be identified, measures will be taken via Entergy's Appendix B corrective action program to provide reasonable confidence that the cause of the condition is determined and corrective action taken to preclude repetition. In Reference 1, Entergy stated that this requested alternative will not impact/change the existing RI-ISI program.

In Reference 3, Entergy states that the guidance for the treatment of repair/replacement activities will be maintained in a high-level corporate document or procedure. The treatment will include guidance such as, but not limited to, design controls, procurement, installation, inspection, and configuration control. Entergy also clarified via an email dated February 16, 2009 (Reference 18), that as part of the alternative repair/replacement treatment, the original construction code fracture toughness requirements will be met.

The NRC staff reviewed the elements of the owner-defined repair/replacement treatment guidance specified by Entergy in Reference 3. The proposed guidance for repair/replacement activities will provide that items will meet the original construction code fracture toughness requirements along with procedural guidance for the applicable processes (design controls, procurement, installation, inspection, and configuration control). The NRC staff finds that these owner-defined controls for repair/replacement activities, if effectively implemented along with the ANO-2 RI-ISI program, will provide an acceptable level of treatment for the LSS SSCs to

- 13 provide reasonable confidence that the repaired or replaced items will remain capable of performing their safety function(s) under design-basis conditions and are consistent with the requirements in 10 CFR 50.69.

4.0 COMMITMENTS In its letter dated February 20, 2008 (Reference 3), the licensee made the following commitments:

1.

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.

2.

The analyses performed in support of R&R activities will follow the requirement in the proposed methodology for crediting operator action.

The scheduled completion dates for both commitments are "Upon implementation of AN02 R&R- 004, Revision 1."

The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the regulatory commitments are best provided by the licensee's administrative processes, including its commitment management program. The regulatory commitments do not warrant the creation of regulatory requirements (items requiring prior NRC approval of subsequent changes).

5.0 CONCLUSION

S The NRC staff has found that the licensee's proposal to use a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items or their associated supports (exclusive of Class CC and MC items) is acceptable.

The licensee's categorization method consists, in essence, of the consequence assessment portion of its NRC-approved RI-ISI methodology, supplemented with the "additional considerations" contained in its RI-RRA method. As described above, the NRC staff has reviewed the differences between the previously endorsed passive SSC categorization N-660 and the licensee's RI-RRA method and concluded that all substantive changes retained in the final RI-RRA methodology are acceptable. The NRC staff has also reviewed the licensee's discussion of the adequacy of its PRA and associated analyses to support the categorization and determined that the quality is sufficient to support assigning piping system failures consequences into broad safety-significant categories. Therefore, the NRC staff concludes that Entergy's process will appropriately identify LSS segments for which repair and replacement activities may be modified as described above.

Based on the information provided in the licensee's submittal, the NRC staff concludes that the licensee's proposed alternative AN02-R&R-004, Revision 1 to the requirements of ASME Code,

- 14 Section XI, IWA-4000 is acceptable because it will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the alternative is authorized for ANO-2 for the remainder of the third and the fourth 1O-year ISI interval.

This authorization for the use of the ANO-2 proposed alternative Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems does not obviate the requirements of 50.55a(g)(4)(ii) for updating the lSI and Repair/Replacement Programs to the latest edition and addenda referenced in 50.55a(b)(2) for the fourth 10-year interval. All other requirements of the ASME Code,Section XI, for which relief has not been specifically requested remain applicable, including a third-party review by the Authorized Nuclear Inservice Inspector.

6.0 REFERENCES

1.

J. F. McCann, Entergy Operations, Inc., letter CI\\IRO-2007-00015, to U.S. l\\Iuclear Regulatory Commission, "Request for Alternative AN02-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," dated April 17, 2007 (ADAMS Accession No. ML071150108).

2.

J. F. McCann, Entergy Operations, Inc., letter CNRO-2007-00028, to U.S. Nuclear Regulatory Commission, "Request for Alternative AN02-R&R-004, Revision 1 Response to NRC Request for Additional Information," dated August 7,2007 (ADAMS Accession No. ML072220160).

3.

D. E. James, Entergy Operations, Inc., letter 2CAN020804, to U.S. Nuclear Regulatory Commission, "Responses to Request for Additional Information to Request for Alternative AN02-R&R-004, Revision 1," dated February 20, 2008 (ADAMS Accession No. ML080520186).

4.

D. E. James, Entergy Operations, Inc., letter 2CAN01 0901, to U.S. Nuclear Regulatory Commission, "Supplemental Information Associated With Treatment of Non-Safety Significant Systems Associated with the Request for Alternative AN02-R&R-004, Revision 1," dated January 12, 2009 (ADAMS Accession 1\\10. ML090120620).

5.

J. Hannon, U.S. Nuclear Regulatory Commission, letter to C. Hutchinson, Entergy Operations, Inc., "Request to Use a Risk-Informed Alternative to the Requirements of ASME Code Section XI, Table IWX-2500 at Arkansas Nuclear One, Unit NO.2 (TAC No.

M99756)," dated December 29, 1998 (ADAMS Legacy Library Accession No.

9901050347).

6.

Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline,"

Revision 0, July 2005 (ADAMS Accession No. ML052900163).

7.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.201, Revision 1, "Guidelines For Categorizing Structures, Systems, and Components In Nuclear Power Plants

- 15 According To Their Safety Significance, For Trial Use," May 2006 (ADAMS Accession No. ML061090627).

8.

American Society of Mechanical Engineers (ASME) Code Case N-660, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1," July 2002.

9.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," October 2007 (ADAMS Accession No. ML072070419).

10.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002 (ADAMS Accession No. ML023240437).

11.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," January 2007 (ADAMS Accession No. ML070240001 ).

12.

American Society of Mechanical Engineers (ASME) RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum B to ASME RA-S-2002, New York, NY, December 30,2005.

13.

B&W Owners Group, Framatome Cogema Fuels, Inc., and Westinghouse Electric Company, NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision A3, Nuclear Energy Institute, Washington, DC, March 20, 2000 (ADAMS Accession No. ML003728023).

14.

ASME Code Case N-662, "Alternative Repair/Replacement Requirements for Items Classified in Accordance With Risk-Informed Processes,Section IX, Division 1," August 2002.

15.

D. Mims, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request to use a Risk-Informed Alternative to the Requirements of ASME Code Section XI, Table IWX-2500 at Arkansas Nuclear One, Unit No.2," (TAC No. M99756), dated September 30,1997 (ADAMS Legacy Library Accession No. 9710070298).

16.

J. Vandergrift, Entergy Operations, Inc., letter to U.S. Nuclear RegUlatory Commission, "Request to Use a Risk-Informed Alternative to the Requirements of ASME Code Section XI, Table IWX-2500 at Arkansas Nuclear One, Unit NO.2" (TAC No. M99756),

dated October 8, 1998 (ADAMS Legacy Library Accession No. 9810190295).

17.

Electric Power Research Institute (EPRI) TR-112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, Final Report," December 1999 (ADAMS Accession No. ML013470102).

- 16

18.

B. W. Clark, Entergy Operations, Inc., electronic mail, to U.S. Nuclear Regulatory Commission, "Arkansas Nuclear One, Unit 2 - Email Regarding Risk-Informed Treatment, Re: AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment. Repair/Replacement Activities in Class 2 & 3 Moderate Energy Systems (TAC No. MD5250)" dated February 16, 2009 (ADAMS Accession No. ML090480061 ).

Principal Contributors: K. Hoffman S. Dinsmore Date:

April 22, 2009

- 2 If you have any questions regarding the SE, please contact Alan B. Wang at (301) 415-1445.

Sincerely, IRA!

Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-368

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:

PUBLIC LPLIV R/F RidsAcrsAcnw_MailCTR Resource RidsNrrDciCpnb Resource RidsNrrDorlLpl4 Resource RidsNrrDraApla Resource RidsNrrLAJBurkhardt Resource RidsNrrPMANO Resource RidsOgcRp Resource RidsRgn4MailCenter Resource S. Dinsmore, NRR/DRNAPLA K. Hoffman, NRR/DCI/CPNB EGuthrie, EDO RIV ADAMS Accession No'.. ML090930246

• SE Input Memo OFFICE NRR/LPL4/PE NRR/LPL4/PM NRR/LPL4/LA DRAIAPLAlBC (A)

DCI/CNPB/BC OGC NRR/LPL4/BC NAME NDiFrancesco AWang JBurkhardt AHowe" TChan" BHarris MMarkley DATE 4/7109 4/7109 4/6/09 3/31/09 3/31/09 4/16/09 4/22/09 OFFICIAL RECORD COpy