NRC-10-0004, Submittal of the Inservice Inspection / Nondestructive Examination Program Relief Requests for the Third Ten-year Interval: Difference between revisions

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{{#Wiki_filter:Joseph H. Plona Site Vice President 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.5910 Fax: 734.586.4172 DTE Energy*
10 CFR 50.55a January 20, 2010 NRC-10-0004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington DC 20555-0001
 
==Reference:==
Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
 
==Subject:==
Submittal of the Inservice Inspection / Nondestructive Examination Program Relief Requests for the Third Ten-year Interval Pursuant to 10 CFR 50.55a(a)(3), Detroit Edison hereby requests NRC approval of the following relief requests for the third ten-year interval of the Inservice Inspection (ISI) /
Nondestructive Examination (NDE) program at Fermi 2 which started on May 2, 2009:
* RR-A30, Continuation of Risk-Informed Inservice Inspection (RI-ISI) application on circumferential welds in Class 1 piping
* RR-A36, Alternative Pressure Testing Requirements for the RPV Flange Leak-off piping
* RR-A37, Alternative Requirements for Examination of Boiling Water Reactor (BWR) Nozzle Inner Radius Sections and Nozzle-to-Shell Welds
* RR-A38, Implementation of Appendix VIII, Supplements 4 and 6 - Use of PDI Qualified Procedures, Personnel and Equipment for Non-Appendix VIII RPV Shell-to-Flange Weld and Head-to-Flange Weld The enclosure to this letter provides details of these relief requests.
Detroit Edison requests NRC approval of these relief requests by September 30, 2010 to support planned testing for the third ten-year ISI/NDE Interval during the next refuel outage scheduled in the fall of 2010.
 
USNRC NRC-10-0004 Page 2 Should you have any questions or require additional information, please contact Mr. Rodney W. Johnson of my staff at (734) 586-5076.
Sincerely, Enclosure cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 4, Region III Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission
 
Enclosure to NRC-10-0004 Fermi 2 Docket No. 50-341 NRC License No. NPF-43 ISI/NDE Relief Requests for Third Ten Year Interval
 
Enclosure to                          10 CFR 50.55a Request NRC-10-0004                              Number RR-A30 Page 1                                Third Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
                  -Alternative Provides Acceptable Level of Quality and Safety-
: 1. ASME Code Components Affected ASME Code Class:            Code Class 1
 
==Reference:==
ASME Section XI, Table IWB-2500-1 Examination Categories:      B-F and B-J Item Numbers:                B5.10, B5.20 and B9.11
 
== Description:==
Continuation of Risk-Informed Inservice Inspection (RI-ISI) application on circumferential welds in Class 1 piping Components:                  All non-exempt Class 1 circumferential piping welds
: 2. Applicable Code Edition and Addenda ASME Section XI, 2001 Edition through 2003 Addenda
: 3. Applicable Code Requirement Class 1 circumferential piping welds are subject to volumetric and surface examinations as stipulated in ASME Section XI, Table IWB-2500-1, Examination Categories B-F and Category B-J.
: 4. Reason for Request The continued use of a risk-informed process as an alternative for the selection of Class 1 Piping Welds for examination is requested for the Third Interval of Fermi 2.
: 5. Proposed Alternative and Basis for Use As an alternative to the Code Requirement, a Risk-Informed process will continue to be used for selection of Class 1 Piping Welds for examination.
The Fermi 2 ISI program for the examination of Class 1 piping welds is currently in accordance with a risk-informed process developed based on EPRI TR-112657, Revision B-A, with identified differences, and with additional guidance taken from ASME Code Case N-578. On April 30, 2001, the Fermi 2 Nuclear Power Plant submitted ISI Relief Request RR-A30 to the NRC, requesting relief from the ASME Section XI Code examination requirements of Class 1 weld (Examination Categories B-F and B-J) inservice inspections
 
Enclosure to                          10 CFR 50.55a Request NRC-10-0004                              Number RR-A30 Page 2                                Third Interval Relief Request (Continued) by implementing a Risk-Informed Inservice Inspection (RI-ISI) Program. Relief Request RR-A30 was approved by the NRC in a letter dated September 10, 2001. Both the original RI-ISI submittal and the resultant NRC Safety Evaluation call for a periodic review and update. To satisfy the periodic review requirements, an evaluation and update was performed in accordance with the Nuclear Energy Institute document 04-05, "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems", published in April, 2004.
In accordance with NEI 04-05, the following aspects were considered during the review:
* Plant Examination Results
* Piping Failures
          -Plant Specific Failures
          -Industry Failures
* PRA Updates
* Plant Design Changes
          -Physical Changes
          -Programmatic Changes
          -Procedural Changes
* Changes in Postulated Conditions
          -Physical Conditions
          -Programmatic Conditions The updated program resulting from this review is the subject of this proposed alternative.
In accordance with the guidance provided by NEI 04-05, Table 1 is provided identifying the number of welds added to and deleted from the originally approved RI-ISI program. The changes to the original program are attributable to three specific actions:
: 1) Condition Assessment Resolution Document (CARD) 07-25329 evaluated limitations of weld 102-304A and determined that it was not inspectable. There was no other weld in Jet Pump Instrumentation (JPI) that was a suitable replacement, so an additional weld in Reactor Coolant Recirculation (RCR) with the same degradation mechanism and a higher Risk Category (2-303J) was selected as a replacement. This caused the total for JPI Risk Category 4(2) to decrease by one and the total for RCR Risk Category 2(2) to increase by one.
: 2) Condition Assessment Resolution Document (CARD) 07-26347 evaluated limitations of weld 4-303A and determined that it was not inspectable. It was replaced by weld 2-203A, which has the same degradation mechanism, but a higher risk. This caused the total for RCR Risk Category 4(2) to decrease by one and the total for RCR Risk Category 2(2) to increase by one.
 
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: 3) Condition Assessment Resolution Document (CARD) 07-26900 reported that a first time Risk-Informed Inservice Inspection (RI-ISI) weld selection was scheduled for examination (FW-RD-2-A1-W1, 4" sweepolet to cap). The examiner reported the scan would be limited due to cap configuration and could not be credited as a full examination for the RI-ISI Program. This weld was not examined, and another selection was required. There were no other 4" selections that were not also limited.
The action taken to correct the condition was to select two adjacent single side access welds (SW-RS-2-A3-W4 and SW-RS-2-A3-W5) to replace the original 4" selection in the Recirculation system. The new selections have the same degradation mechanism and consequence of failure as the original welds. CARD 07-26900 also reported that a weld selection previously examined (FW-RD-B1-W1) was likely impacted by current expectations for weld flatness. It would be limited due to cap configuration and should not be credited as a full examination for the RI-ISI Program.
Another selection was required. There were no other 4" selections that were not also limited. The action taken to correct the condition was to select two adjacent single side access welds (SW-RS-2-B3-W4 and SW-RS-2-B3-W5) to replace the original 4" selection in the Recirculation system. The new selections have the same degradation mechanism and consequence of failure as the original welds. These actions caused the total for RCR Risk Category 4(2) to increase by two.
The actions listed above are the only ones resulting in changes to the number of welds selected for examination per the RI-ISI Program. There were a limited number of additional "like-for-like" substitutions for weld selections that were made during the implementation of the program due to inaccessibility and ALARA. However, these actions had no impact on the total numbers of elements inspected.
A new Risk Impact Analysis was performed, and the revised program continues to satisfy the acceptance criteria of Regulatory Guide 1.174 and EPRI TR-112657 when compared to the last deterministic Section XI inspection program.
The Risk-Informed process continues to provide an adequate level of quality and safety for selection of the Class 1 Piping Welds for examination. Therefore, pursuant to 10CFR50.55a(a)(3)(i), it is requested that the proposed alternative be authorized.
: 6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019, subject to the review and update guidance of NEI 04-05.
 
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: 7. Precedent Fermi 2 Second Interval Relief Request RR-A30, "Risk-Informed Inspection Program Plan Fermi Nuclear Power Plant, Unit 2", as approved by the NRC in a letter dated September 10, 2001 (ADAMS Accession No. ML012400331).
: 8. PRA Quality The current PRA model (FermiV7) addresses internal events (including internal flooding) at full power. The model incorporates recent advances in PRA technology across many elements. These elements include the proper characterization of initiating events involving Loss of Offsite Power (LOOP), treatment of time dependant offsite power recovery, treatment of operator actions to implement bus ties and other Emergency Operating Procedures (EOPs), equipment success criteria calculations, data analysis of key parameters, maintenance unavailability, and common cause failure probabilities.
In the Level 2 analysis, Containment Fault Trees (CFTs) are developed to provide the link between the plant damage states associated with core damage mitigation and containment integrity with the possible radionuclide releases of varying timing and magnitudes. The model considers the performance of the reactor building and sprays in the assessment of radionuclide mitigation. The spectrum of radionuclide releases that could result from the core damage condition is then calculated for the postulated discrete end states of the CFT.
The CFTs model the various potential radionuclide release paths to the environment and provide an estimate of their relative likelihoods. This process is an iterative one, requiring technical feedback between the system fault trees, the CFTs, and the plant response evaluation. The purpose of containment fault trees is to provide estimates of the conditional probabilities of various radionuclide releases and timing given the core damage sequences defined in the Fermi Level 1 PRA. The Large Early Release Frequency (LERF) is represented by one of these linked CFTs. This approach to the LERF evaluation also supports realistic quantification of systematic contributions to containment isolation failures (bypass sequences that are actually linked to the Level 1 model).
PRA Self Assessment and Peer Review Several assessments of technical capability have been made, and continue to be planned, for the Fermi 2 PRA model. These assessments are further discussed in the paragraphs below.
* An independent PRA Peer Review was conducted under the auspices of the BWR Owners Group in 1997, following the Industry PRA Peer Review process. This peer review included an assessment of the PRA model maintenance and update process.
* During 2005 and 2006 the Fermi 2 PRA model results were evaluated in the BWR Owners Group PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process.
 
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* In 2008, a gap analysis was performed against the ASME PRA Standard and the Regulatory Guide 1.200 Revision 1.
* Another gap analysis will be performed prior to the RG 1.200 Peer Review to reflect pertinent changes to both the PRA Standard and Regulatory Guide 1.200.
* An independent PRA Peer Review of the Fermi 2 PRA model against RG 1.200 Revision 1 is planned for Spring 2012.
A summary of the disposition of 1997 Industry PRA Peer Review Facts and Observations (F&Os) for the Fermi 2 PRA models was documented as part of the statement of PRA capability for MSPI in the Fermi 2 MSPI Basis Document. As noted in that document, there were no significance level A or B F&Os which were open at the time of the review (with the exception of 4 documentation related B level F&Os, which did not adversely affect the MSPI applications). It was noted also in the MSPI submittal there were no MSPI cross-comparison outliers for Fermi 2.
A Gap Analysis for the PRA model (version FermiV7) was completed in August 2008. This Gap Analysis was performed against RG 1.200 Revision 1 and the associated ASME Standard. This gap analysis defined a list of 67 supporting requirements from the Standard for which gaps to Capability Category II of the Standard were identified. An independent PRA Peer Review of the Fermi 2 PRA model against RG 1.200 Revision 1 is planned for Spring 2012. An assessment of how each of these gaps affects the RI-ISI application is presented in Attachment 1. Based upon the information in Attachment 1, none of the gaps are deemed significant with respect to the RI-ISI program.
General Conclusion Regarding PRA Capability The Fermi 2 PRA maintenance and update processes and technical capability evaluations described above provide a basis for concluding that the PRA is suitable for use in certain risk-informed licensing actions. As specific risk-informed PRA applications are performed, remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.
Assessment of PRA Capability Needed for Risk-Informed Inservice Inspection (RI-ISI)
In the Risk-Informed Inservice Inspection (RI-ISI) program at Fermi, the EPRI Risk informed ISI methodology is used to define alternative ISI requirements. Plant-specific PRA-derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking, element selection and delta risk evaluation steps.
The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three fundamental elements of the EPRI methodology.
 
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: 1) PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RI-ISI inspection as illustrated below. Broad ranges are used to define these bins so that the impact of uncertainty is minimized and only substantial PRA changes would be expected to have an impact on the consequence ranking results.
Consequence Results Binning Groups Consequence Category                CCDP Range                    CLERP Range High                      CCDP > 1E-4                    CLERP > 1E-5 Medium                  1E-6 < CCDP < 1E-4            1E-7 < CLERP < 1E-5 Low                      CCDP < 1E-6                    CLERP < 1E-7 The risk importance of a weld is therefore not tied directly to a specific PRA result.
Instead, it depends only on the range in which the PRA result falls. As a consequence, any PRA modeling uncertainties would be mitigated by the wide binning provided in the methodology. Additionally, conservatism in the binning process (e.g., as would typically be introduced through PRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to result in a larger inspection population.
: 2) The impacts of particular PRA consequence results are further dampened by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment. The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment (and ultimately each element) according to the EPRI Risk Matrix. The Risk Matrix, which equally takes both assessments into consideration, is reproduced below.
 
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CONSEQUENCES OF PIPE RUPTURE POTENTIAL FOR                IMPACTS ON CONDITIONAL CORE DAMAGE PROBABILITY PIPE RUPTURE                        AND LARGE EARLY RELEASE PROBABILITY PER DEGRADATION MECHANISM SCREENING CRITERIA NONE            LOW        MEDIUM          HIGH HIGH                  LOW          MEDIUM          HIGH          HIGH FLOW ACCELERATED CORROSION    Category 7    Category 5    Categoir 3    Category I MEDIUM                  LOW            LOW        MEDIUM          HIGH OTHER DEGRADATION MECHANISMS    Category 7    Cgory 6 C6    Category 5    C      y2 LOW                  LOW            LOW          LOW          MEDIUM NO DEGRADATION MECHANISMS      Category 7    Category 7    Category 6    Category 4
: 2) The EPRI RI-ISI methodology uses an absolute risk ranking approach. As such, conservatism in either the consequence assessment or the failure potential assessment will result in a larger inspection population rather than masking other important components. That is, providing more realism into the PRA model (e.g., by meeting higher capability categories) most likely would result in a smaller inspection population.
These three elements of the methodology reduce the importance and influence of PRA on the final list of candidate welds.
The limited manner of PRA involvement in the RI-ISI process is also reflected in the risk-informed license application guidance provided in Regulatory Guide 1.174. Section 2.2.6 of Regulatory Guide 1.174 provides the following insight into PRA capability requirements for this type of application:
There are, however, some applications that, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.
An example is risk-informed inservice inspection (RI-ISI). In this application, risk significance was used as one criterion for selecting pipe segments to be periodically examined for cracking. During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary. Therefore, the staff review of
 
Enclosure to                        10 CFR 50.55a Request NRC-10-0004                              Number RR-A30 Page 8                                Third Interval Relief Request (Continued) plant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability.
Conclusion Regarding PRA Capability for Risk-Informed ISI The Fermi 2 PRA model continues to be suitable for use in the RI-ISI application. This conclusion is based on:
* The technical adequacy of the PRA is appropriate for the application.
* The PRA technical capability evaluations that have been performed and which are scheduled for the future.
* The RI-ISI process considerations, as noted above, that demonstrate the relatively limited sensitivity of the EPRI RI-ISI process to PRA attribute capability beyond ASME PRA Standard Capability Category I.
 
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Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category SysteRisk            Consequence        Failure Potential      Code    Weld    Original            Revised System        Category      Rank        Rank          DMs            Rank    Category Count R      OtherRI-ISI R          Otther(1 RPV              6        Low      Medium          None            Low      B-J    3      0                    0 CRD            4 (2)    Mdium          High      None (IGSCC) Low (Medium)    B-F      1    1(2)                1(2)
(High)
Medium                                                B-F      2    1(3 )              0(11)
M e(Hig        High      None (IGSCC) Low (Medium)      B-      2    01                  0 JPI            4 (2)
(High)                                                B-J      2    0                    0 MS              4      Medium        High          None            Low      B-J    105    11                  11 MS              6        Low        Medium          None            Low      B-J      8    0                    0 High (High)    High      CC, (IGSCC)      MediuB-F            10    3                  5( 4 12 )
RCR            2 (2)
(Medium)
Medium                                                B-F      2    1(5)                1(5)
RCR            4(2)      (High        High      None (IGSCC) Low (Medium)      B-      71    76                8(6 11, 2 (High)                                                B-J    71 RCR              4      Medium        High          None            Low      B-J    38    4                    4 Medium                                                B-F      3    1(7)                1(7)
RHR            4 (2)      (High        High      None (IGSCC) Low (Medium)      B-      3    0                    0 (High)                                                B-J      3    0                    0 RHR              4      Medium        High          None            Low      B-J    22    3                    3 RHR              6        Low        Medium          None            Low      B-J    46    0                    0
 
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Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk            Consequence        Failure Potential    Code      Weld      Original    Revised System        Category        Rank        Rank          DMs            Rank  Category Count RI-ISI Other(l' RI-ISI Other(l Medium                      1(8)        1(8)
CS            2 (2)    High  (High)
Medium        High      CC, (IGSCC)      (Medium M(9)      B-F      2      1) 1((9)      1)
CS            4 (2)      Mdium          High    None (IGSCC) Low (Medium)    B-F      2      1)
(High)                                                    I        I      I_I_I CS              4        Medium        High          None            Low      B-J      24      3          3 CS              6          Low        Medium          None            Low      B-J      18      0          0 Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk                                Failure Potential              Weld      Original    Revised d  Count Sym        Consequence                            Category        t System    Category        Rank        Rank          Rank              ank  RSI Other RI-ISI Other'1' RI-ISI Other HPCI              4        Medium        High          None            Low    B-J      12        2          2 HPCI              6          Low        Medium          None            Low    B-J      2        0          0 RCIC              4        Medium        High          None            Low    B-J      14        2          2 RCIC              6          Low        Medium          None            Low    B-J      2        0          0
 
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Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk                                      Failure Potential                Weld      Original      Revised d  Count Consequence                                  Category System      Category          Rank Rank            Rank R an k          DMs DMs          Rank Rank      RI-ISI    (2)
                                                                                            -SI  Other  RI-ISI Other(1 RI-ISI Other(1)
Medium                                                      B-F B-        2 6      0
: 10)            0 1()
RWCU            4 (2)        Medim            High        None (IGSCC) Low (Medium)
RWCU              4          Medium          High            None          Low        B-J      53      6            6 RWCU              6            Low          Medium            None          Low        B-J      11      0            0 FW              2            High            High        TASCS, CC        Medium        B-J      12      6            6 FW              2            High            High            TASCS        Medium        B-J      23      3            3 FW              4          Medium          High            None          Low        B-J      82      9            9 FW              6            Low          Medium            None          Low        B-J      6      0            0 Notes are shown in parenthesis:
: 1. The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-112657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, Fermi 2 achieved greater than a 10% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.
 
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Notes for Table 1 (cont'd):
: 2. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
: 3. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
: 4. These three welds were selected for examination by both the IGSCC Program and the RI-ISI Program. Since crevice corrosion was identified along with IGSCC as a potential damage mechanism for these welds, the examinations will include the requirements identified in EPRI TR-112657 for crevice corrosion examinations in order to be credited toward both the IGSCC and the RI-ISI Programs.
: 5. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
: 6. These welds were selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for these welds, the IGSCC examinations will be credited toward both programs.
: 7. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
: 8. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since crevice corrosion was identified along with IGSCC as a potential damage mechanism for this weld, the examination will include the requirements identified in EPRI TR-112657 for crevice corrosion examinations in order to be credited toward both the IGSCC and the RI-ISI Programs.
 
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: 9. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
10.This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
11 .A weld previously selected was determined to have limitations. An alternate weld was selected.
12.Additional single side access welds were selected to compensate for dual side welds that could not be examined due to configuration.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 14                                          Third Interval Relief Request (Continued) 1    Document basis for exclusion of system alignments        IE-A4a    Not Met  None. This gap is a result of the need to (human induced failures). Refer to consensus                                document the exclusion of human induced report EPRI 1013492 Table 5-3; Human induced                                failures. It is not anticipated that modeling Initiators are included in the generic priors used                          changes will be necessary to address this from NUREG/CR-5750.                                                        SR.
2    Perform a review of plant data (e.g. CARDs, Daily        IE-A7    Met (I) Not Significant. Precursor review is Work request) for the purpose of identifying                                unlikely to significantly alter initiators additional initiating events,                                              relevant to RI-ISI. This SR was ranked as a CC I. This is in keeping with the ranking required for RI-ISI.
3    Update CAFTA model with updated IE frequencies          IE-Cla    Not Met  Not significant. A review of the updated IE that were generated using the Bayes methodology,                            frequencies revealed little change to those initiators relevant to RI-ISI. The consequence rank for the initiators will not change from that in the FermiV7 model.
4    SR IE-C10 directs to compare and explain                IE-C10    Not Met  Not Significant. Current IE frequency differences in IE frequency results with generic                            development is consistent with industry sources. Such documentation is not provided in the                          standards. A cross comparison with other EF2 IE Notebook. Provide a comparison of EF2                                data sources is primarily a concern for frequencies to industry generic sources in the                              documentation only.
Summary section of the EF2 IE Notebook (or in an appendix). One can use NUREG/CR-5750 values.
 
Enclosure to                                        10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 15                                              Third Interval Relief Request (Continued) 5    Develop a final report (or database) which can be          IE-D1    Not Met  None. This is a documentation issue. It is referenced to demonstrate that each of the ASME                              not anticipated that the model will be elements has been considered.                                                modified to address this item.
6    Complete the "Uncertainty Analysis" section                IE-D3    Not Met  Not Applicable. The EPRI RI-ISI process is associated with each section of the PRA analysis,                            structured such that model uncertainties will The EF2 PSA model-of-record and associated                                    not unduly influence results. Also, Draft documentation need to be enhanced to include                                  EPRI Report, "PRA Technical Adequacy sensitivity assessments of key assumptions and                              Guidance for Risk-Informed Inservice parameters (e.g., the list of sensitivities                                  Inspection Programs." states that this incorporated into the BWROG Option 2 Guidelines                              requirement need not be met for RI-ISI are generically applicable).                                                applications.
7    Develop a final report (or database) which can be        AS-C1      Not Met None. See Gap #5.
referenced to demonstrate that each of the ASME elements has been considered.
 
Enclosure to                                      10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 16                                            Third Interval Relief Request (Continued) 8    Documentation requires updating and corrections to      AS-C2      Not Met  None. See Gap #5.
be better aligned with latest model. It contains outdated information. For example (but not limited to):
: 1. Suppression pool must be initiated by 4 hours not 6 as stated in the document.
: 2. Documentation refers to several Event Trees (ET) where the current model has only one ET.
: 3. Discussion in 4.2.3 regarding RISKMAN model may only lead to confusion. It may be better to remove references to RISKMAN Event Trees.
: 4. Documentation indicates a 6 hour short-term HPCI/RCIC success, but the current model utilizes a 4 hour mission time.
9    Add Accident Sequence Analysis section to the            AS-C3      Not Met Not Applicable. See Gap #6.
        "Uncertainty Analysis" section associated with the each section of the PRA analysis.
10    Update mission time for exceptions to the 24 hour        SC-A5      Met (I) None. This SR was ranked as a CC I. This limit to be internally consistent (e.g., HPCI/RCIC 4                        is in keeping with the ranking required for versus 6 hours). No other actions are required.                            RI-ISI in Draft EPRI Report, "PRA Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 17                                            Third Interval Relief Request (Continued)
PRA S(amlard to RI-ISI Sulmfttal Ap,licAbflitv of Weitified Gips to Capability CategorN (CC) ll of ASMI1 11    Some reasonableness checks have been performed            SC-B5    Not Met  Not Significant. Due to the conservatism (MSPI system success Criteria calculation results                          inherent in the EPRI RI-ISI methodology, are comparable to MAAP calculation results),                                the completion of reasonableness checks on Complete reasonableness check and acceptability of                          all evaluations is judged to be non-the results of the thermal/hydraulic, structural, or                        significant. It should be noted that there other supporting engineering bases used to support                          were no significant deviations identified for the success criteria.                                                      the calculations where reasonableness checks have already been performed.
12    Develop a final report (or database) which can be        SC-C1    Not Met  None. See Gap #5.
referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models.
Follow their lead and prepare a Fermi specific document Consider creating success criteria sections similar to MSPI systems for all systems/ functions modeled in the PRA.
13    Complete the "Uncertainty Analysis" section              SC-C3    Not Met Not Applicable. See Gap #6.
associated with the each section of the PRA analysis.
 
Enclosure to                                      10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 18                                            Third Interval Relief Request (Continued) 14    Review in detail with system engineers and                SY-A4    Not Met Not Significant. It is not anticipated that operators to confirm that the systems analysis                              this process will have a significant impact correctly reflects the as-built, as-operated plant.                        on the quantification results with respect to Document this review.                                                      RI-ISI.
15    The PRA model is judged to include proper              SY-A12      Not Met Not Significant. Due to the conservatism treatment of components and failure modes for                              inherent in the EPRI RI-ISI methodology, Capability Category II requirements. Additional                            the lack of documentation of a search for investigation in determining whether all appropriate                        additional equipment and failure modes is components and failure modes are included could                            judged to be non-significant.
be performed; however, this is judged not to have significant beneficial impact on the model.
To fully address this SR it is necessary to document a search for additional component and failure modes.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 19                                            Third Interval Relief Request (Continued)
Attachnent I 16    The PRA model is judged to include proper                SY-A14  Not Met Not Significant. It is not anticipated that treatment of components and failure modes for                              this process will have a significant impact Capability Category II requirements. Additional                            on the quantification results with respect to investigation in determining whether all appropriate                      RI-ISI.
components and failure modes are included could be performed; however, this is judged not to have significant beneficial impact on the model.
In order to explicitly meet this SR it would be necessary to provide documentation of why components or failure modes are not included in the model utilizing the screening criteria cited by this SR.
17    The current HVAC dependency analysis uses                SY-B 13  Not Met Not Significant. The level of detail subjective judgment or operator action as a basis for                      provided in the subjective HVAC excluding modeling HVAC for certain areas. This                            dependency analysis cited is adequate to was appropriate for the IPE analysis but this                              address the RI-ISI application.
analysis should be improved to meet the ASME standard.
 
Enclosure to                                      10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 20                                            Third Interval Relief Request (Continued) 18    Develop a final report (or database) which can be      SY-C1      Not Met None. See Gap #5.
referenced to demonstrate that each of the ASME elements has been considered.
There is no documented independent review of system notebooks. Develop system notebooks and have them reviewed.
19    Many system notebooks do not address several            SY-C2    Not Met None. See Gap #5.
ASME SRs (i.e., System Fault trees, evidence of system engineer review etc.).
Improve documentation by explicitly addressing as many of the supporting requirements as possible.
Document interviews with system engineers and plant operators that confirm that the system analysis reflects the as-built, as-operated plant.
Update dependency matrix utilizing CAFTA nomenclature. Also revise to include HVAC support and other systems included in the PRA model.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                RR-A30 Page 21                                          Third Interval Relief Request (Continued)
A,ttachinent 20    Complete the "Uncertainty Analysis" section            SY-C3    Not Met  Not Applicable. See Gap #6.
associated with each section of the PRA analysis.
21    Update the review of test and maintenance              HR-B1    Not Met  Not significant. Pre-initiator human actions procedures. Document the additional rules utilized                        are judged to not significantly impact RI-ISI for screening individual activities from further                          applications.
consideration.
22    Document review of activities that may cause            HR-C2      Met (I) None. This SR was ranked as a CC I. This failure of automatic realignment (e.g., EDG start                          is in keeping with the ranking required for signals). Include a review of other failure modes                          RI-ISI in Draft EPRI Report, "PRA identified during update of the unavailability                            Technical Adequacy Guidance for Risk-analysis                                                                  Informed Inservice Inspection Programs."
 
Enclosure to                                      10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 22                                            Third Interval Relief Request (Continued) 23    A review of HRA documentation has uncovered            HR-E3      Not Met  Not Significant. The Top 30 risk significant some discrepancies regarding the referenced                                operator actions (by RAW) have been procedures (CARD 07-21694). Talk through (i.e.                              reviewed by the operating and training staff.
review in detail) with plant operators to confirm                          The remainder of the human actions are of that the HR analysis correctly reflects the as-                            low significance.
operated plant. Document this review.
Establish a method to monitor procedure changes.                            Not Applicable. This gap relates to capturing a method to ensure procedures are updated with respect to their impact on the HRA. This is a program maintenance and documentation issue which does not affect the current RI-ISI submittal.
24    Significant operator actions should be reviewed in      HR-E4      Met (I) Not Significant. The Top 30 risk significant detail by the operating staff and their impact                              operator actions (by RAW) have been included in the HRA evaluation. Use simulator                              reviewed by the operating staff. The observations or talk-through with operators to                            remainder of the human actions are of low confirm the response for scenarios modeled.                                significance.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 23                                            Third Interval Relief Request (Continued) 25    Significant operator actions should be reviewed in      HR-G5      Met (I) None. This SR was ranked as a CC I. This detail by the operating staff and the associated                            is in keeping with the ranking required for comments included in the HRA evaluation. Use of                            RI-ISI in Draft EPRI Report, "PRA simulator observations or talk-through with                                Technical Adequacy Guidance for Risk-operators to confirm the action times for scenarios                        Informed Inservice Inspection Programs."
modeled.
26    Develop a final report (or database) which can be        HR-I1    Not Met  None. See Gap #5.
referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models.
Follow their lead and prepare a Fermi specific document.
27    Document the processes used to identify,                HR-I2    Not Met  None. See Gap #5.
characterize and quantify the pre-initiator, post-initiator and recovery actions considered in the PRA (including the inputs, methods, and results).
 
Enclosure to                                        10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 24                                              Third Interval Relief Request (Continued) 28    Complete the "Uncertainty Analysis" section                HR-I3    Not Met  Not Applicable. See Gap #6.
associated with each section of the PRA analysis.
The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).
29    There is no discussion provided on the type of            DA-A2      Not Met  Not significant. This is primarily a distribution utilized. Add discussion on this issue                          documentation issue. The distributions to the PSA documentation,                                                    utilized at Fermi are generally consistent with best practices for the failure modes modeled.
30    In order to meet Category II, parameter estimation        DA-B1      Met (I) None. This SR was ranked as a CC I. This must group components based on service condition                              is in keeping with the ranking required for (e.g., clean vs. untreated water, air).                                      RI-ISI in Draft EPRI Report, "PRA Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."
31    Provide evidence that this "interview" process has        DA-C12      Met (I) None. This SR was ranked as a CC I. This and continues to occur, and/or perform interviews                            is in keeping with the ranking required for with plant maintenance and operation staff to                                RI-ISI in Draft EPRI Report, "PRA ensure a reliable estimate of unavailability for                            Technical Adequacy Guidance for Risk-significant basic events.                                                    Informed Inservice Inspection Programs."
 
Enclosure to                                      10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 25                                            Third Interval Relief Request (Continued)
Ga .c...ri.tono                o f (a)                      SR  ,p    cc .nc.            1,  or. to R 1-IST 32    The approach suggested is to identify dominant risk      DA-C13    Not Met    Not significant. The model is reasonably contributor combinations based on knowledge of                                consistent with known plant operating the accident sequence modeling, and model such                                practice and experience. An exhaustive combinations of coincident maintenance outages in                              assessment is not needed to support the use the fault tree logic. A review of recent maintenance                          of PRA for RI-ISI.
experience should then be performed to identify events of coincident maintenance outages for these combinations to support probability estimation for the events 33    There is no evidence that this check was made.            DA-D4    Met (I)  None. This SR was ranked as a CC I. This Check that the posterior distribution is reasonable                          is in keeping with the ranking required for given the relative weight of evidence provided by                            RI-ISI in Draft EPRI Report, "PRA the prior and the plant-specific data.                                        Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."
34    Data analysis section when compared to other              DA-E1    Not Met  None. See Gap #5.
elements is the most complete but it is still in draft form. Develop a final report (or database) which can be referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models. Follow their lead and prepare a Fermi specific document.
 
Enclosure to                                        10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 26                                              Third Interval Relief Request (Continued) 35    Complete the "Uncertainty Analysis" section              DA-E3      Not Met  Not Applicable. See Gap #6.
associated with each section of the PRA analysis.
The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).
36    Document the screening process and results to            IF-Bib    Not Met  Not Applicable. The internal flooding PRA reduce the scenarios which require further analysis.                          model was not used to support the development of the Fermi RI-ISI Program for Class I piping.
37    To meet strict reading of SR IF-B3, the                    IF-B3    Not Met Not Applicable. See Gap #36 characteristics (e.g., pressure and temperature, etc.)
of the various flooding sources and mechanisms need to be documented.
38    PERFORM necessary engineering calculations for            IF-C3c    Not Met Not Applicable. See Gap #36 flood rate, time to reach susceptible equipment, and the structural capacity of SSCs in accordance with the applicable requirements described in Table 4.5.3-2(b) of the ASME standard.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 27                                            Third Interval Relief Request (Continued)
Ajflicibility (41dewified Gaps to Ca          ifitN Catc,ory (CC) 11 of AS11I PWX Stamljird to RI-ISt Sulmittal 39    Update IE frequency utilizing the applicable              IF-D5    Not Met  Not Applicable. See Gap #36 requirements in Table 4.5.1-2 of the ASME standard. Update flood initiator frequency utilizing updated generic sources, also include uncertainty values.
40    Perform and document the scenario screening                IF-D7    Not Met  Not Applicable. See Gap #36 process.
41    Screen out a flood area if the product of the sum of      IF-E3a    Not Met  Not Applicable. See Gap #36 the frequencies of the flood scenarios for the area, and the bounding conditional core damage probability (CCDP) is less than 10-9/reactor year.
The bounding CCDP is the highest of the CCDP values for the flood scenarios in an area.
42    The internal flooding HEP calculations need to be          IF-E5    Not Met  Not Applicable. See Gap #36 enhanced to be consistent with the requirements of HLR HR. These HEP calculations should be integrated into the HRA notebook and performed using the EPRI HRA Calculator.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 28                                            Third Interval Relief Request (Continued) 43    There are no HRA Calculator entries related to          IF-E5a    Not Met  Not Applicable. See Gap #36 flood scenarios. Conservative estimates are used without details on how they were estimated (e.g.
A3G10-SM-2).
Two types of HRAs should be evaluated.
: 1. Review HRA values that are in the CAFTA model that would be impacted by the flood scenarios. If necessary copy and modify values to include flood initiator impact.
: 2. Enter necessary data into HRA Calculator for flood specific HRAs. These HEP calculations should be integrated into the HRA Notebook.
44    Only six flood initiators are included in the Fermi      IF-E6    Not Met Not Applicable. See Gap #36 fault tree model. Other initiators should be added if they do not meet the flood screening criteria. The number that should be added is dependent on completing the flood screening process.
45    Perform review of LERF calculations for                  IF-E7    Not Met Not Applicable. See Gap #36 unscreened flood scenarios.
 
Enclosure to                                        10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 29                                              Third Interval Relief Request (Continued)
Attacimeit I 46    Develop a final report (or database) which can be          IF-F1    Not Met None. See Gap #5.
referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models.
Follow their lead and prepare a Fermi specific document. Update documentation to reflect CAFTA model nomenclature.
47    Complete the "Uncertainty Analysis" section                IF-F3    Not Met Not Applicable. See Gap #6.
associated with each section of the PRA analysis.
The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).
 
Enclosure to                                        10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 30                                              Third Interval Relief Request (Continued) 48    Document the review of a sampling of non-                QU-D4      Not Met  Not Significant. Due to the conservatism significant accident cutsets to ensure they are                              inherent in the EPRI RI-ISI methodology, reasonable and have physical meaning.                                        the lack of documentation of a review of non-significant accident sequences is judged to be negligible. It should be noted that an aggregate review of all accident sequences was performed prior to release of the current model.
49    Provide documentation that includes identification        QU-D5a      Met (I) None. This SR was ranked as a CC I. This of significant common cause failure contributors to                          is in keeping with the ranking required for CDF. To attain CC II, initiator fault trees would                            RI-ISI in Draft EPRI Report, "PRA have to be developed quantified to present                                    Technical Adequacy Guidance for Risk-importance values.                                                            Informed Inservice Inspection Programs."
50    Complete the "Uncertainty Analysis" section              QU-El      Not Met Not Applicable. See Gap #6.
associated with each section of the PRA analysis.
The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).
 
Enclosure to                                        10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 31                                              Third Interval Relief Request (Continued)
GLII -                    Des,cripoio of Gap                      SR              I,,  1m$ortanc              to RMSt 51    Complete the "Uncertainty Analysis" section              QU-E2      Not Met    Not Applicable. See Gap #6.
associated with each section of the PRA analysis.
The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).
52    To be consistent with SR QU-E4, the EF2 PRA                QU-E4      Met (I)    None. This SR was ranked as a CC I. This model and associated documentation should be                                    is in keeping with the ranking required for enhanced. The enhancement should include                                        RI-ISI in Draft EPRI Report, "PRA sensitivity assessments of the impact of uncertain                              Technical Adequacy Guidance for Risk-model boundary conditions and logical                                            Informed Inservice Inspection Programs."
combinations of uncertainties.
The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                RR-A30 Page 32                                          Third Interval Relief Request (Continued) plicabilitN                _11s t " ability Cae!gor    C)11 of ASI11PRA S(anaad to RI-ISI Subtit 53    Develop a final report (or database) which can be      QU-F1    Not Met  None. See Gap #5.
referenced to demonstrate that each of the ASME elements has been considered.
 
Enclosure to                                      10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 33                                            Third Interval Relief Request (Continued) 54    Strict reading of SR QU-F1 would indicate that the        QU-F2    Not Met None. See Gap #5.
following enhancements to the documentation of the EF2 PRA would need to be made to comply with the Standard:
(c) Add a general description of the quantification process including accounting for systems successes.
(d) Document the process and results for establishing the truncation screening values for final quantification demonstrating that convergence towards a stable result was achieved.
(g) Add a discussion on equipment or human actions that are the key factors in causing the accidents to be non-dominant.
(h) Perform and document all sensitivity studies.
(i) Update the uncertainty distribution for the total CDF (3-33)
(1) Document asymmetries in quantitative modeling to provide application users the necessary understanding regarding why such asymmetries are present in the model.
 
Enclosure to                                        10 CFR 50.55a Request Number NRC-10-0004                                                    RR-A30 Page 34                                              Third Interval Relief Request (Continued) 55    Complete the "Uncertainty Analysis" section                QU-F4    Not Met  Not Applicable. See Gap #6.
associated with each section of the PRA analysis.
The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).
56    Document limitations in the quantification process        QU-F5    Not Met  None. See Gap #5.
that would impact applications 57    DEVELOP system models that support the accident            LE-C5      Not Met Not Significant. This gap is a result of progression analysis consistent with the applicable                            some of the Level 2 support systems not requirements for paragraph 4.5.4, as appropriate for                          being modeled and/or documented to the the level of detail of the analysis.                                          same standards as the Level 1 model. It is judged that the level of system modeling or alternate approximations utilized at Fermi are adequate to support the RI-ISI application.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 35                                            Third Interval Relief Request (Continued) 58    Some of the Level 2 HRA analysis is not performed        LE-C6    Not Met  Not significant. This is primarily a to the same level of detail as Level 1 HRAs.                                documentation quality issue for the Level 2 HRA basic events. The EPRI methodology Add the same level of details to the Level 2 HRAs                          bounds many of the uncertainties inherent in as included in Level 1 HRAs. Ensure updated                                the Level 2 modeling.
procedures are utilized for this HRA analysis.
Adjust HRA values as appropriate.
59    Review significant LERF sequences and evaluate if      LE-C9b      Met (I) None. This SR was ranked as a CC I. This an engineering analysis can support equipment                              is in keeping with the ranking required for continued operation or operator action after                                RI-ISI in Draft EPRI Report, "PRA containment failure.                                                        Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."
60    Review significant LERF sequences and evaluate if      LE-C10      Met (I) None. This SR was ranked as a CC I. This an engineering analysis can support equipment                              is in keeping with the ranking required for continued operation or operator action after                                RI-ISI in Draft EPRI Report, "PRA containment failure.                                                      Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."
61    PERFORM a realistic interfacing system failure          LE-D3      Met (I) None. This SR was ranked as a CC I. This probability analysis for the significant accident                            is in keeping with the ranking required for progression sequences resulting in a large early                            RI-ISI in Draft EPRI Report, "PRA release.                                                                    Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                RR-A30 Page 36                                          Third Interval Relief Request (Continued)
ApliRcahilit' ot'We'ified G-41n to CapabilitY CaygorN (CC) II ol'AS-Nl  PZA Stanarid to 10-IS1 SOu      nittal 62    Document all Level 2 equipment failures and            LE-E1    Not Met None. See Gap #5.
operator actions consistent with Level 1 requirements.
63    Strict reading of SR LE-E4 would indicate that the    LE-E4    Not Met Not significant. Do to the conservative following enhancements to the LERF analysis and                          nature of the RI-ISI process, Level 2 HEP associated documentation would need to be made to                        dependencies are judged not to be comply with the Standard:                                                significant. The EPRI methodology bounds
* Explicitly assess dependencies among Level 2                          many of the uncertainties inherent in the HEPs (and combinations of Level 2 HEPs with                        Level 2 modeling.
Level 1 HEPs)
* Complete quantitative uncertainty assessment of the LERF analysis (refer to Disposition for SR LE-F2). Sensitivity runs should be updated.
 
Enclosure to                                    10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 37                                            Third Interval Relief Request (Continued) 64    Lack of traceability from current CAFTA model to        LE-G1    Not Met  None. Additional documentation and model Level 2 (IPE) documentation. Level 2 analysis                              nomenclature changes are needed to close refers to the large event trees used with the                              this gap. It is not anticipated that the model RISKMAN calculation. Currently the CAFTA                                    nomenclature changes will affect the model is a large fault tree model. Documentation                            quantification results.
should address this difference in methodology and be revised accordingly.
: 1. Rename Level 2 Basic Event to be consistent with Level 1 BEs.
: 2. Construct a translation matrix between the CAFTA Model and the Level 2 (IPE) Report.
: 3. Create a CAFTA Level 2 Report.
65    A review of Work Request (WR) database which            MU-B2      Not Met Not Applicable. The prioritization of open tracks model changes has inconsistent priority (1-4)                        model work requests in the PSA Work given to significance (A-D) ratings.                                      Tracking Database (WTDB) has no impact Establish a consistent method for assigning priority                        on the results used to support the RI-ISI based on significance. Ensure significance ofWRs                            program.
is properly assessed. Re-evaluate WR in the database and assign higher priority to model changes that have potential of having the highest impact on the core damage risk.
 
Enclosure to                                  10 CFR 50.55a Request Number NRC-10-0004                                                  RR-A30 Page 38                                          Third Interval Relief Request (Continued)
Attcadinient I ApL)ficab)ifity ot hicuified Gaps to Captbflit8 CateioiN (CC) 11 of ASall PRA Stanlair] to RI-ISI Sulmittal 66  The EF2 PRA received a PRA peer review in 1997          MU-B4      Not Met None. The Self-Assessment (Gap Analysis) using the BWROG peer review guidelines. The                                process is being used to define the current PRA was converted from the RISKMAN software                                Fermi PRA technical adequacy for the RI-platform. A new peer review should be performed                            ISI submittal. A Peer Review is currently to address both 1) the change in software platform,                        scheduled for the Spring of 2010.
and 2) the more stringent PRA peer review requirements per the ASME PRA Standard (Addendum B).
67  The PRA maintenance and update procedures                MU-C1    Not Met Not Applicable. The lack of explicit should include guidelines for the cumulative impact                        guidelines for assessing the cumulative of pending changes on risk applications. This can                          impact of pending changes on risk be done during the WR evaluation process.                                  applications does not impact the RI-ISI application. It should be noted that the RI-ISI results are reviewed after every major model release.
Notes:
: 1. SR: Supporting Requirement
: 2. SR Codes are contained in ASME RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," December 2005.
 
Enclosure to                          10 CFR 50.55a Request NRC-10-0004                              Number RR-A36 Page 39                                Third Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)
                                  -Hardship or Unusual Difficulty without a Compensating    Increase in Level of Quality or Safety-
: 1. ASME Code Components Affected ASME Code Class:            Code Class 1
 
==References:==
ASME Section XI, Table IWB-2500-1 and IWB-5222 Examination Category:      B-P (All Pressure Retaining Components)
Item Number:                B15.10
 
== Description:==
Alternative Pressure Testing Requirements for the RPV Flange Leak-Off Piping Components:                NPS 1 RPV Flange Seal Leak-Off Piping
 
===2. Applicable Code Edition and Addenda===
ASME Section XI, 2001 Edition through 2003 Addenda
 
===3. Applicable Code Requirement===
IWB-2500, Table IWB-2500-1, Code Category B-P, Item Number B15.10 requires that all Class 1 pressure retaining components be Visual, VT-2 examined each refueling outage. The required system pressure test can be either a hydrostatic test or a system leakage test. The system leakage test is performed at a pressure not less than the pressure corresponding to 100% rated reactor power. Per IWB-5222(a), the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second closed valve at the boundary extremity. Per IWB-5222(b), the pressure retaining boundary during the system leakage test conducted at or near the end of the interval shall extend to all Class 1 pressure retaining components within the system boundary.
 
Enclosure to                          10 CFR 50.55a Request NRC-10-0004                              Number RR-A36 Page 40                                Third Interval Relief Request (Continued)
 
===4. Reason for Request===
As discussed in 3, "Applicable Code Requirements", ASME Section XI, 2001 Edition through 2003 Addenda requires that Class 1 pressure boundary piping shall be pressure tested after each refueling outage. The Reactor Pressure Vessel (RPV) head flange seal leak detection piping is separated from the reactor coolant pressure boundary by one passive membrane, which is an O-ring located on the inner vessel flange as shown in Attachment 1.
A second O-ring is located on the outside of the tap in the vessel flange. Failure of the inner O-ring is the only condition under which this line is pressurized. Therefore, the line is not expected to be pressurized during the system pressure test following a refueling outage.
The configuration of this piping precludes system pressure testing while the vessel head is removed because the configuration of the vessel tap coupled with the high test pressure prevents the tap in the flange from being temporarily plugged or connected to other piping.
The opening in the flange is smooth walled, making the effectiveness of a temporary seal very limited. Failure of a temporary test seal could possibly cause ejection of the device used for plugging or connecting to the vessel flange. To perform the system leakage test in accordance with the Code requirements, the RPV head flange seal detection piping would have to be redesigned, fabricated, and installed. This would impose a severe and unnecessary burden.
The configuration also precludes pressurizing the line with the head installed because the seal prevents complete filling of the piping, which has no vent available. The top head of the vessel contains two grooves that hold the O-rings. The O-rings are held in place by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test were to be performed with the head on, the inner O-ring would be pressurized in a direction opposite to its design function. This test pressure would result in a net inward force on the inner O-ring that would tend to push it into the recessed cavity that houses the retainer clips.
The thin O-ring material would very likely be damaged by the inward force. The design of this line makes the ASME Code required system leakage test impractical either with the vessel head installed or removed.
: 5. Proposed Alternative and Basis for Use In lieu of the requirements of IWB-5222(b), a VT-2 visual examination will be performed on the subject piping during vessel flood-up each refueling outage. The hydrostatic head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross indications in the piping.
The flange seal leak-offline is essentially a leakage collection and detection system. The line would only function as a Class 1 pressure boundary if the inner O-ring fails thereby pressurizing the line. During this time, the control room annunciator would be in alarm. If the annunciator ceases to be in alarm, this would indicate that the outer O-ring or seal leak-off line had failed and resulted in a reactor coolant pressure boundary leak. This would
 
Enclosure to                        10 CFR 50.55a Request NRC-10-0004                              Number RR-A36 Page 41                                Third Interval Relief Request (Continued) require immediate plant shutdown. Per IWA-5243, "when leakage from components are normally expected and collected (such as valve stems, pump seals, or vessel flange gaskets) the visual examination VT-2 shall be conducted by verifying that the leakage collection system is operative."
Fermi 2 has implemented a periodic Preventive Maintenance Event (PM Event B564) to pressurize an isolable section of the leak detection line to verify that the pressure switch is operative and in calibration. Performance of this event satisfies the intent of IWA-5243 for leakage collection systems.
This line is also inspected during the VT-2 system leakage test. There has been no report of gross structural defect to date. Additionally, the line is filled at static head pressure during the outage while the reactor cavity is flooded. Any gross leakage would be easily detected during drywell entries. The VT-2 is considered to be met by PM Event B564 along with a visual inspection of the line during flood-up.
Pursuant to 10 CFR 50.55a(a)(3)(ii), approval is requested to use the proposed alternative described above in lieu of the Code requirement on the basis that the Code requirements establish a hardship or unusual difficulty without a compensating increase in the level of quality or safety.
: 6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019.
: 7. Precedents Fermi 2 Relief Request RR-A36, "Evaluation of Second 10-Year Interval Inservice Inspection Request for Relief No. RR-A36 on end of Interval System Pressure Test" (TAC No. ME0868) as approved by the NRC in a letter dated August 28, 2009.
 
Enclosure to          10 CFR 50.55a Request NRC-10-0004              Number RR-A36 Page 42                Third Interval Relief Request Attachment 1 Reactor Pressure Vessel Seal Leak-Off Details Detroit Edison Fermi 2
 
Enclosure to            10 CFR 50.55 a Request NRC-10-0004                Number RR-A36 Page 43                  Third Interval Relief Request Attachment 1 (Continued)
N    See Figure 2 for detail REATORPRESUR VESELHEAD FLANGE LEAK-OFF LINE CONFIGURATION
 
Enclosure to                      10 CFR 50.55a Request NRC-10-0004                            Number RR-A36 Page 44                            Third Interval Relief Request Attachment 1 (Continued)
Outer Flange      High Pressure Leak Detection Monitoring Tap Seal Ring-, L
                              -    Inner Flange Seal Ring 4      See Detail "A" Detail W*"
Vessel Flange Sectional View Figure 2 REACTOR PRESSURE VESSEL HEAD FLANGE LEAK-OFF LINE DETAILS
 
Enclosure to                10 CFR 50.55a Request Number NRC-10-0004                              RR-A37 Page 45                      Third Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
                -Alternative Provides Acceptable Level of Quality and Safety-
: 1. ASME Code Components Affected ASME Code Class:          Code Class 1
 
==References:==
ASME Section XI, 2001 Edition with 2003 Addenda Code Case N-702 BWRVIP-108: BWR Vessel and Internals Project, "Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1003557, October 2002 Examination Category:      B-D Item Numbers:              B3.90, B3.100
 
== Description:==
Alternative Requirements for Examination of Boiling Water Reactor (BWR) Nozzle Inner Radius Sections and Nozzle-to-Shell Welds Components:                N1, N3, N5, N6, N7, N8, and N10 Nozzles (see Attachment 2 for specific nozzle identifications)
 
===2. Applicable Code Edition and Addenda===
ASME Section XI, 2001 Edition through 2003 Addenda
 
===3. Applicable Code Requirement===
Table IWB-2500-1, Examination Category B-D, Inspection Program B, Item Numbers B3.90, and B3.100 require a volumetric examination to be performed once per interval on each reactor vessel nozzle-to-vessel weld and nozzle inside radius section.
 
Enclosure to                      10 CFR 50.55a Request Number NRC-10-0004                                      RR-A37 Page 46                            Third Interval Relief Request (Continued)
 
===4. Reason for Request===
Leverage the technical basis and criteria in BWRVIP-108 to realize substantial radiation dose and cost savings while maintaining an acceptable level of quality and safety.
: 5. Proposed Alternative and Basis for Relief Pursuant to 10CFR50.55a(a)(3)(i), Detroit Edison requests authorization to utilize the alternative requirements described herein in lieu of the requirements of Table IWB-2500-1, Examination Category B-D, Item Numbers B3.90, and B3.100 for reactor vessel nozzle-to-shell welds and nozzle inside radius sections. These alternative requirements will not be utilized for the feedwater nozzles or the recirculation inlet nozzles. The alternative requirements will be applied to the control rod drive return nozzle which has been cut and capped at Fermi 2 and therefore is not subject to the thermal fatigue issues which might otherwise be a concern for operational control rod drive return lines.
This alternative allows a 25% sampling of the reactor vessel nozzle inner radius section examinations and nozzle-to-shell weld examinations to be implemented, provided at least one nozzle from each system and nominal pipe size is examined. This alternative also allows a VT-1 examination of the nozzle inner radius base metal surfaces to be performed in lieu of the Code required volumetric examination. For the nozzle-to-shell welds requiring examination, a volumetric examination will be performed. ASME Section XI, Appendix VIII, 2001 Edition with no Addenda will be used for volumetric examinations as mandated in 10CFR50.55a(b)(2)(xv).
Current requirements for the inspection of the BWR nozzle-to-shell welds and nozzle blend radii are costly and result in significant radiation exposure to examiners. The performance of NDE has improved substantially since the examinations of ASME Section XI, Table IWB-2500-1, Examination Category B-D, Item Numbers B3.90, and B3.100 were first required such that there is now a high reliability of detecting flaws that can challenge the structural integrity of BWR nozzles and their associated welds. Knowledge of improved NDE capabilities, coupled with fracture mechanics, provides a technical basis to justify reduction of inspections while maintaining safety. This technical basis is provided in BWR Vessel and Internals Project (BWRVIP) Report No. BWRVIP-108. For any cracks in the nozzle blend radius region, the results of BWRVIP-108 show that the conditional failure probability of the nozzles due to a low temperature overpressure (LTOP) event are very low (<1x10-6 for 40 years), even without any inservice inspection. At the nozzle-to-vessel shell weld, the conditional probability of failure due to the LTOP event is also very small (<1x10-6 for 40 years), with or without any inservice inspection. As such, the BWRVIP-108 report provides the technical basis for the reduction of the nozzle-to-shell welds and nozzle blend radii from 100% to 25% of the nozzles every 10 years.
This EPRI report received an NRC SER dated December 19, 2007. In the SER, Section 5.0 "Plant Specific Applicability" indicates that each licensee who plans to request relief from the ASME Code, Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle
 
Enclosure to                        10 CFR 50.55a Request Number NRC-10-0004                                      RR-A37 Page 47                                Third Interval Relief Request (Continued) inner radius sections may reference BWRVIP-108 report as the technical basis for the use of the alternatives presented herein. However, each licensee should demonstrate the plant-specific applicability for the BWRVIP-108 report to its units in the relief request by showing that all the general and nozzle-specific criteria addressed below are satisfied (reference Attachment 1):
(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 115 0 F per hour.
: a. Per TS SR 3.4.10.1, the RPV heatup/cooldown is limited to less than or equal to 100 0F in any 1 hour period. This criterion is met.
(2) For the Recirculation Inlet nozzles, the following criteria must be met:
: a.  (pr/t)/CRpv<l.15, the calculation for the Fermi 2 N2 Nozzle results in 0.89 which is less than 1.15.
: b. [p(ro2+ri2)/(ro2-ri2)]/CNOzzLE< 1.15, the calculation for the Fermi 2 N2 Nozzles results in 1.229 which is greater than 1.15, therefore this criteria is not met for the Recirculation Inlet Nozzles.
(3) For the Recirculation Outlet nozzles, the following criteria must be met:
: a. (pr/t)/CRpv<l.15, the calculation for the Fermi 2 N2 Nozzles results in 1.07 which is less than 1.15.
: b. [p(ro2+ri2)/(r0o2ri2)]/CNOzzLE <1.15, the calculation for the Fermi 2 N2 Nozzles results in 1.069 which is less than 1.15.
Based upon the above information, all RPV nozzle-to-vessel shell welds and nozzle inner radius sections, with the exception of the Recirculation Inlet Nozzles, meet the criteria and therefore are applicable. Since the Recirculation Inlet Nozzles do not meet the criteria, the alternative requirements will not be applied to those nozzles. Additionally, as stated earlier, this alternative will not be applied to the feedwater nozzles.
 
Enclosure to                      10 CFR 50.55a Request Number NRC-10-0004                                    RR-A37 Page 48                            Third Interval Relief Request (Continued)
: 6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019.
: 7. Precedents Duane Arnold Energy Center - Safety Evaluation for Request for Alternative to Reactor Pressure Vessel Nozzle to Vessel Weld and Inner Radius Examinations was approved on August 29, 2008 (Accession number ML080710428, TAC No. MD8193).
 
Enclosure to                            10 CFR 50.55a Request NRC-10-0079                                  Number RR-A37 Page 49                              Third Interval Relief Request (Continued)
Attachment 1 Response to NRC Plant Specific Applicability to BWRVIP-108
: 1. The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 11i5T/hour Response: TS SR 3.4. 10.1 limits the RPV heatup/cooldown to *1&#xfd;l00'F/hour for Curve B and
    *!&#xfd;20 0 F/hour for Curve A.
Recirculation Inlet Nozzles                                Recirculation Outlet Nozzles 2    (pr/t)/CR~pv<1 .15                                    4  (pr/t)/CRPV<1. .15 p=RPV Normal Operating                    1045            p=RLPV Normal Operating                  1045 Pressure                                                  Pressure r--RPV inner radius                        127.125        r-=RPV inner radius                      127.125 t=RPV wall thickness                      7.6875          t=~RPV wall thickness                    7.6875 CRPV=                                      19332          CRPV=                                    16171 (Pr/t)/CRPV=  0.89                                          (pr/t)/CRPV= 1.07 3    [p(r,,2+ri 2 )/(r,,2 -ri 2)]/CNOZZLE<1  .15            5    [P(rol+r, 2)/(ro2 _r, 2)]/CNOZZLE<. .15 p=RPV Normal Operating                    1045            p=RPV Normal Operating                    1045 Pressure                                                    Pressure r,==nozzle outer radius                    11I            ro=nozzle outer radius                    22.5625 ri=nozzle inner radius                    6.1875          rj==nozzle inner radius                  13.125 CNOZZLE                                  1637            CNOZ'ZLE                                1977
[p(ro +ri2 )/(r. 2-ri 2)]/CNOZZLE=        1.229            [p(r. 2 +ri 2)/(r. 2 _ri 2 )]/CNOZZLE=  1.069
 
Enclosure to                            10 CFR 50.55a Request Number NRC-10-0004                                          RR-A37 Page 50                                  Third Interval Relief Request (Continued)
Attachment 2 Applicable Components Code    Item Nozzle Category Number  Identification          Description                System      Isometric N3    B-D      B3.100 8-316A IRS        Nozzle Inside Radius Section Main Steam          5361-5 N3    B-D      B3.100 8-316B IRS        Nozzle Inside Radius Section Main Steam          5361-5 N3    B-D      B3.100 8-316C IRS        Nozzle Inside Radius Section Main Steam          5361-5 N3    B-D      B3.100 8-316D IRS        Nozzle Inside Radius Section Main Steam          5361-5 N5    B-D      B3.100 14-316A IRS      Nozzle Inside Radius Section Core Spray          5361-5 N5    B-D      B3.100 14-316B IRS      Nozzle Inside Radius Section Core Spray          5361-5 N10  B-D      B3.100 15-315 IRS        Nozzle Inside Radius Section Control Rod Drive  5361-5 Reactor Recirc N1    B-D      B3.100 5-314A IRS        Nozzle Inside Radius Section Suction            5361-5 Reactor Recirc N1    B-D      B3.100 5-314B IRS        Nozzle Inside Radius Section Suction            5361-5 Jet Pump N8    B-D      B3.100 19-314A IRS      Nozzle Inside Radius Section Instrument          5361-5 Jet Pump N8    B-D      B3.100 19-314B IRS      Nozzle Inside Radius Section Instrument          5361-5 N7    B-D      B3.100 2-318 IRS        Nozzle Inside Radius Section RPV Head Vent      5361-5 N6    B-D      B3.100 4-318A IRS        Nozzle Inside Radius Section RPV Head Spare      5361-5 N6  B-D      B3.100 4-318B IRS        Nozzle Inside Radius Section RPV Head Spare      5361-5 N3  B-D      B3.90  8-316A            M.S. Nozzle-to-Vessel Weld  Main Steam          5361-5 N3  B-D      B3.90  8-316B            M.S. Nozzle-to-Vessel Weld  Main Steam          5361-5 N3  B-D      B3.90  8-316C            M.S. Nozzle-to-Vessel Weld  Main Steam          5361-5 N3  B-D      B3.90  8-316D            M.S. Nozzle-to-Vessel Weld  Main Steam          5361-5 N5  B-D      B3.90  14-316A          C.S. Nozzle-to-Vessel Weld  Core Spray          5361-5 N5  B-D      B3.90  14-316B          C.S. Nozzle-to-Vessel Weld  Core Spray          5361-5 CRD Ref. Nozzle-to-Vessel N10  B-D      B3.90  15-315            Weld                        Control Rod Drive  5361-5
 
Enclosure to                          10 CFR 50.55a Request Number RR-A37 NRC-10-0004                                  Third Interval Relief Request Page 51                                                (Continued)
Attachment 2 Applicable Components (continued)
Code    Item Nozzle Category Number  Identification          Description            System    Isometric Reactor Recirc.
N1    B-D      B3.90  5-314A          RRI Nozzle-to-Vessel Weld  Suction          5361-5 Reactor Recirc.
N1    B-D      B3.90  5-314B          RRI Nozzle-to-Vessel Weld  Suction          5361-5 Jet Pump N8    B-D      B3.90  19-314A          JPI Nozzle-to-Vessel Weld  Instrument        5361-5 Jet Pump N8    B-D      B3.90  19-314B          JPI Nozzle-to-Vessel Weld  Instrument        5361-5 Head/Vent Nozzle-to-Vessel N7    B-D      B3.90  2-318            Weld                      RPV Head Vent    5361-5 Spare Nozzle-to-Vessel N6  B-D      B3.90  4-318A            Weld                      RPV Head Spare    5361-5 Spare Nozzle-to-Vessel N6  B-D      B3.90  4-318B            Weld                      RPV Head Spare    5361-5
 
Enclosure to                  10 CFR 50.55a Request Number NRC-10-0004                                RR-A38 Page 52                        Third  Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
                -Alternative Provides Acceptable Level of Quality and Safety-
: 1. ASME Code Components Affected ASME Code Class:          Code Class 1
 
==References:==
ASME Section XI, Table IWB-2500-1 ASME Section XI, Appendix VIII, Supplements 4 and 6, 10CFR50.55a(b)(xv).
Examination Category:      B-A Item Numbers :            B1.30 and B1.40
 
== Description:==
Implementation of Appendix VIII, Supplements 4 and 6 - Use of PDI Qualified Procedures, Personnel and Equipment for Non-Appendix VIII RPV Shell-to-Flange Weld and Head-to-Flange Weld.
Components:                Reactor Pressure Vessel (RPV) Shell-to-Flange Weld No. 13-308 Reactor Pressure Vessel (RPV) Head-to-Flange Weld No. 3-319
 
===2. Applicable Code Edition and Addenda===
ASME Section XI, 2001 Edition (No Addenda) per 10CFR50.55a(b)(2)(xv) for Appendix VIII.
 
===3. Applicable Code Requirement===
The 2001 Edition with No Addenda of the American Society of Mechanical Engineers (ASME Code) Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Subsection IWA-2232, requires ultrasonic (UT) examinations be performed in accordance with Mandatory Appendix I. Paragraph I-2110(b) of Mandatory Appendix I requires that examination of the RPV shell-to-flange weld and head-to-flange weld to be in accordance with ASME Code, Section V, Article 4.
 
Enclosure to                    10 CFR 50.55a Request Number NRC-10-0004                                  RR-A38 Page 53                          Third Interval Relief Request (Continued)
 
===4. Reason for Request===
The use of this alternative will allow the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel RPV flange welds from the accessible vessel or head side of the welds in accordance with ASME Code, Section XI, Division 1, 2001 Edition, No Addenda, Appendix VIII, Supplements 4 and 6. This alternative would be used in lieu of Article 4 of Section V requirements.
: 5. Proposed Alternative and Basis for Relief Detroit Edison proposes to perform ultrasonic examinations of the RPV shell-to-flange weld and head-to-flange weld using procedures, personnel, and equipment that have been demonstrated and qualified in accordance with ASME Section XI, 2001 Edition, No Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDI Program. Since the examinations will be performed from a single side due to the weld configuration, all procedures, personnel, and equipment will be qualified for single sided access for examination of these welds.
Appendix VIII requirements were developed and adopted to ensure the effectiveness of ultrasonic examinations within the nuclear industry by means of a rigorous, item specific performance demonstration containing flaws of various sizes, locations, and orientations.
The performance demonstration process has established with a high degree of confidence, the capability of personnel, procedures, and equipment to detect and characterize flaws that could be detrimental to the structural integrity of the RPV. The PDI approach has demonstrated that for detection and characterization of flaws in the RPV the ultrasonic examination techniques are equal to, or surpass the requirements of the ASME Section V, Article 4 ultrasonic examination requirements. Though Appendix VIII is not specified for the RPV flange weld UT examinations, the use of Appendix VIII, Supplements 4 and 6 criteria for detection and sizing of flaws in these welds will be equal to, or exceed the requirements of ASME Section V, Article 4. Therefore, the use of the proposed alternative will continue to provide an acceptable level of quality and safety, and approval is requested pursuant to 10 CFR 50.55a(a)(3)(i).
 
Enclosure to                    10 CFR 50.55a Request Number NRC-10-0004                                  RR-A38 Page 54                          Third Interval Relief Request (Continued)
: 6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019.
: 7. Precedents A similar relief request (RR ISI-30) was previously approved for Union Electric Company for its Callaway Plant, Unit 1 on April 7, 2004 (ADAMS Accession Nos. ML032340608 and ML041000516).}}

Revision as of 22:33, 13 November 2019

Submittal of the Inservice Inspection / Nondestructive Examination Program Relief Requests for the Third Ten-year Interval
ML100220171
Person / Time
Site: Fermi DTE Energy icon.png
Issue date: 01/20/2010
From: Plona J
DTE Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-10-0004
Download: ML100220171 (57)
v • d • e


Text

Joseph H. Plona Site Vice President 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.5910 Fax: 734.586.4172 DTE Energy*

10 CFR 50.55a January 20, 2010 NRC-10-0004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington DC 20555-0001

Reference:

Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

Subject:

Submittal of the Inservice Inspection / Nondestructive Examination Program Relief Requests for the Third Ten-year Interval Pursuant to 10 CFR 50.55a(a)(3), Detroit Edison hereby requests NRC approval of the following relief requests for the third ten-year interval of the Inservice Inspection (ISI) /

Nondestructive Examination (NDE) program at Fermi 2 which started on May 2, 2009:

  • RR-A30, Continuation of Risk-Informed Inservice Inspection (RI-ISI) application on circumferential welds in Class 1 piping
  • RR-A36, Alternative Pressure Testing Requirements for the RPV Flange Leak-off piping
  • RR-A37, Alternative Requirements for Examination of Boiling Water Reactor (BWR) Nozzle Inner Radius Sections and Nozzle-to-Shell Welds
  • RR-A38, Implementation of Appendix VIII, Supplements 4 and 6 - Use of PDI Qualified Procedures, Personnel and Equipment for Non-Appendix VIII RPV Shell-to-Flange Weld and Head-to-Flange Weld The enclosure to this letter provides details of these relief requests.

Detroit Edison requests NRC approval of these relief requests by September 30, 2010 to support planned testing for the third ten-year ISI/NDE Interval during the next refuel outage scheduled in the fall of 2010.

USNRC NRC-10-0004 Page 2 Should you have any questions or require additional information, please contact Mr. Rodney W. Johnson of my staff at (734) 586-5076.

Sincerely, Enclosure cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 4, Region III Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission

Enclosure to NRC-10-0004 Fermi 2 Docket No. 50-341 NRC License No. NPF-43 ISI/NDE Relief Requests for Third Ten Year Interval

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A30 Page 1 Third Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Components Affected ASME Code Class: Code Class 1

Reference:

ASME Section XI, Table IWB-2500-1 Examination Categories: B-F and B-J Item Numbers: B5.10, B5.20 and B9.11

Description:

Continuation of Risk-Informed Inservice Inspection (RI-ISI) application on circumferential welds in Class 1 piping Components: All non-exempt Class 1 circumferential piping welds

2. Applicable Code Edition and Addenda ASME Section XI, 2001 Edition through 2003 Addenda
3. Applicable Code Requirement Class 1 circumferential piping welds are subject to volumetric and surface examinations as stipulated in ASME Section XI, Table IWB-2500-1, Examination Categories B-F and Category B-J.
4. Reason for Request The continued use of a risk-informed process as an alternative for the selection of Class 1 Piping Welds for examination is requested for the Third Interval of Fermi 2.
5. Proposed Alternative and Basis for Use As an alternative to the Code Requirement, a Risk-Informed process will continue to be used for selection of Class 1 Piping Welds for examination.

The Fermi 2 ISI program for the examination of Class 1 piping welds is currently in accordance with a risk-informed process developed based on EPRI TR-112657, Revision B-A, with identified differences, and with additional guidance taken from ASME Code Case N-578. On April 30, 2001, the Fermi 2 Nuclear Power Plant submitted ISI Relief Request RR-A30 to the NRC, requesting relief from the ASME Section XI Code examination requirements of Class 1 weld (Examination Categories B-F and B-J) inservice inspections

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A30 Page 2 Third Interval Relief Request (Continued) by implementing a Risk-Informed Inservice Inspection (RI-ISI) Program. Relief Request RR-A30 was approved by the NRC in a letter dated September 10, 2001. Both the original RI-ISI submittal and the resultant NRC Safety Evaluation call for a periodic review and update. To satisfy the periodic review requirements, an evaluation and update was performed in accordance with the Nuclear Energy Institute document 04-05, "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems", published in April, 2004.

In accordance with NEI 04-05, the following aspects were considered during the review:

  • Plant Examination Results
  • Piping Failures

-Plant Specific Failures

-Industry Failures

  • Plant Design Changes

-Physical Changes

-Programmatic Changes

-Procedural Changes

  • Changes in Postulated Conditions

-Physical Conditions

-Programmatic Conditions The updated program resulting from this review is the subject of this proposed alternative.

In accordance with the guidance provided by NEI 04-05, Table 1 is provided identifying the number of welds added to and deleted from the originally approved RI-ISI program. The changes to the original program are attributable to three specific actions:

1) Condition Assessment Resolution Document (CARD) 07-25329 evaluated limitations of weld 102-304A and determined that it was not inspectable. There was no other weld in Jet Pump Instrumentation (JPI) that was a suitable replacement, so an additional weld in Reactor Coolant Recirculation (RCR) with the same degradation mechanism and a higher Risk Category (2-303J) was selected as a replacement. This caused the total for JPI Risk Category 4(2) to decrease by one and the total for RCR Risk Category 2(2) to increase by one.
2) Condition Assessment Resolution Document (CARD) 07-26347 evaluated limitations of weld 4-303A and determined that it was not inspectable. It was replaced by weld 2-203A, which has the same degradation mechanism, but a higher risk. This caused the total for RCR Risk Category 4(2) to decrease by one and the total for RCR Risk Category 2(2) to increase by one.

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A30 Page 3 Third Interval Relief Request (Continued)

3) Condition Assessment Resolution Document (CARD) 07-26900 reported that a first time Risk-Informed Inservice Inspection (RI-ISI) weld selection was scheduled for examination (FW-RD-2-A1-W1, 4" sweepolet to cap). The examiner reported the scan would be limited due to cap configuration and could not be credited as a full examination for the RI-ISI Program. This weld was not examined, and another selection was required. There were no other 4" selections that were not also limited.

The action taken to correct the condition was to select two adjacent single side access welds (SW-RS-2-A3-W4 and SW-RS-2-A3-W5) to replace the original 4" selection in the Recirculation system. The new selections have the same degradation mechanism and consequence of failure as the original welds. CARD 07-26900 also reported that a weld selection previously examined (FW-RD-B1-W1) was likely impacted by current expectations for weld flatness. It would be limited due to cap configuration and should not be credited as a full examination for the RI-ISI Program.

Another selection was required. There were no other 4" selections that were not also limited. The action taken to correct the condition was to select two adjacent single side access welds (SW-RS-2-B3-W4 and SW-RS-2-B3-W5) to replace the original 4" selection in the Recirculation system. The new selections have the same degradation mechanism and consequence of failure as the original welds. These actions caused the total for RCR Risk Category 4(2) to increase by two.

The actions listed above are the only ones resulting in changes to the number of welds selected for examination per the RI-ISI Program. There were a limited number of additional "like-for-like" substitutions for weld selections that were made during the implementation of the program due to inaccessibility and ALARA. However, these actions had no impact on the total numbers of elements inspected.

A new Risk Impact Analysis was performed, and the revised program continues to satisfy the acceptance criteria of Regulatory Guide 1.174 and EPRI TR-112657 when compared to the last deterministic Section XI inspection program.

The Risk-Informed process continues to provide an adequate level of quality and safety for selection of the Class 1 Piping Welds for examination. Therefore, pursuant to 10CFR50.55a(a)(3)(i), it is requested that the proposed alternative be authorized.

6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019, subject to the review and update guidance of NEI 04-05.

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A30 Page 4 Third Interval Relief Request (Continued)

7. Precedent Fermi 2 Second Interval Relief Request RR-A30, "Risk-Informed Inspection Program Plan Fermi Nuclear Power Plant, Unit 2", as approved by the NRC in a letter dated September 10, 2001 (ADAMS Accession No. ML012400331).
8. PRA Quality The current PRA model (FermiV7) addresses internal events (including internal flooding) at full power. The model incorporates recent advances in PRA technology across many elements. These elements include the proper characterization of initiating events involving Loss of Offsite Power (LOOP), treatment of time dependant offsite power recovery, treatment of operator actions to implement bus ties and other Emergency Operating Procedures (EOPs), equipment success criteria calculations, data analysis of key parameters, maintenance unavailability, and common cause failure probabilities.

In the Level 2 analysis, Containment Fault Trees (CFTs) are developed to provide the link between the plant damage states associated with core damage mitigation and containment integrity with the possible radionuclide releases of varying timing and magnitudes. The model considers the performance of the reactor building and sprays in the assessment of radionuclide mitigation. The spectrum of radionuclide releases that could result from the core damage condition is then calculated for the postulated discrete end states of the CFT.

The CFTs model the various potential radionuclide release paths to the environment and provide an estimate of their relative likelihoods. This process is an iterative one, requiring technical feedback between the system fault trees, the CFTs, and the plant response evaluation. The purpose of containment fault trees is to provide estimates of the conditional probabilities of various radionuclide releases and timing given the core damage sequences defined in the Fermi Level 1 PRA. The Large Early Release Frequency (LERF) is represented by one of these linked CFTs. This approach to the LERF evaluation also supports realistic quantification of systematic contributions to containment isolation failures (bypass sequences that are actually linked to the Level 1 model).

PRA Self Assessment and Peer Review Several assessments of technical capability have been made, and continue to be planned, for the Fermi 2 PRA model. These assessments are further discussed in the paragraphs below.

  • An independent PRA Peer Review was conducted under the auspices of the BWR Owners Group in 1997, following the Industry PRA Peer Review process. This peer review included an assessment of the PRA model maintenance and update process.
  • During 2005 and 2006 the Fermi 2 PRA model results were evaluated in the BWR Owners Group PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process.

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A30 Page 5 Third Interval Relief Request (Continued)

  • An independent PRA Peer Review of the Fermi 2 PRA model against RG 1.200 Revision 1 is planned for Spring 2012.

A summary of the disposition of 1997 Industry PRA Peer Review Facts and Observations (F&Os) for the Fermi 2 PRA models was documented as part of the statement of PRA capability for MSPI in the Fermi 2 MSPI Basis Document. As noted in that document, there were no significance level A or B F&Os which were open at the time of the review (with the exception of 4 documentation related B level F&Os, which did not adversely affect the MSPI applications). It was noted also in the MSPI submittal there were no MSPI cross-comparison outliers for Fermi 2.

A Gap Analysis for the PRA model (version FermiV7) was completed in August 2008. This Gap Analysis was performed against RG 1.200 Revision 1 and the associated ASME Standard. This gap analysis defined a list of 67 supporting requirements from the Standard for which gaps to Capability Category II of the Standard were identified. An independent PRA Peer Review of the Fermi 2 PRA model against RG 1.200 Revision 1 is planned for Spring 2012. An assessment of how each of these gaps affects the RI-ISI application is presented in Attachment 1. Based upon the information in Attachment 1, none of the gaps are deemed significant with respect to the RI-ISI program.

General Conclusion Regarding PRA Capability The Fermi 2 PRA maintenance and update processes and technical capability evaluations described above provide a basis for concluding that the PRA is suitable for use in certain risk-informed licensing actions. As specific risk-informed PRA applications are performed, remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.

Assessment of PRA Capability Needed for Risk-Informed Inservice Inspection (RI-ISI)

In the Risk-Informed Inservice Inspection (RI-ISI) program at Fermi, the EPRI Risk informed ISI methodology is used to define alternative ISI requirements. Plant-specific PRA-derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking, element selection and delta risk evaluation steps.

The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three fundamental elements of the EPRI methodology.

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1) PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RI-ISI inspection as illustrated below. Broad ranges are used to define these bins so that the impact of uncertainty is minimized and only substantial PRA changes would be expected to have an impact on the consequence ranking results.

Consequence Results Binning Groups Consequence Category CCDP Range CLERP Range High CCDP > 1E-4 CLERP > 1E-5 Medium 1E-6 < CCDP < 1E-4 1E-7 < CLERP < 1E-5 Low CCDP < 1E-6 CLERP < 1E-7 The risk importance of a weld is therefore not tied directly to a specific PRA result.

Instead, it depends only on the range in which the PRA result falls. As a consequence, any PRA modeling uncertainties would be mitigated by the wide binning provided in the methodology. Additionally, conservatism in the binning process (e.g., as would typically be introduced through PRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to result in a larger inspection population.

2) The impacts of particular PRA consequence results are further dampened by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment. The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment (and ultimately each element) according to the EPRI Risk Matrix. The Risk Matrix, which equally takes both assessments into consideration, is reproduced below.

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CONSEQUENCES OF PIPE RUPTURE POTENTIAL FOR IMPACTS ON CONDITIONAL CORE DAMAGE PROBABILITY PIPE RUPTURE AND LARGE EARLY RELEASE PROBABILITY PER DEGRADATION MECHANISM SCREENING CRITERIA NONE LOW MEDIUM HIGH HIGH LOW MEDIUM HIGH HIGH FLOW ACCELERATED CORROSION Category 7 Category 5 Categoir 3 Category I MEDIUM LOW LOW MEDIUM HIGH OTHER DEGRADATION MECHANISMS Category 7 Cgory 6 C6 Category 5 C y2 LOW LOW LOW LOW MEDIUM NO DEGRADATION MECHANISMS Category 7 Category 7 Category 6 Category 4

2) The EPRI RI-ISI methodology uses an absolute risk ranking approach. As such, conservatism in either the consequence assessment or the failure potential assessment will result in a larger inspection population rather than masking other important components. That is, providing more realism into the PRA model (e.g., by meeting higher capability categories) most likely would result in a smaller inspection population.

These three elements of the methodology reduce the importance and influence of PRA on the final list of candidate welds.

The limited manner of PRA involvement in the RI-ISI process is also reflected in the risk-informed license application guidance provided in Regulatory Guide 1.174. Section 2.2.6 of Regulatory Guide 1.174 provides the following insight into PRA capability requirements for this type of application:

There are, however, some applications that, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.

An example is risk-informed inservice inspection (RI-ISI). In this application, risk significance was used as one criterion for selecting pipe segments to be periodically examined for cracking. During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary. Therefore, the staff review of

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A30 Page 8 Third Interval Relief Request (Continued) plant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability.

Conclusion Regarding PRA Capability for Risk-Informed ISI The Fermi 2 PRA model continues to be suitable for use in the RI-ISI application. This conclusion is based on:

  • The technical adequacy of the PRA is appropriate for the application.
  • The PRA technical capability evaluations that have been performed and which are scheduled for the future.
  • The RI-ISI process considerations, as noted above, that demonstrate the relatively limited sensitivity of the EPRI RI-ISI process to PRA attribute capability beyond ASME PRA Standard Capability Category I.

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Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category SysteRisk Consequence Failure Potential Code Weld Original Revised System Category Rank Rank DMs Rank Category Count R OtherRI-ISI R Otther(1 RPV 6 Low Medium None Low B-J 3 0 0 CRD 4 (2) Mdium High None (IGSCC) Low (Medium) B-F 1 1(2) 1(2)

(High)

Medium B-F 2 1(3 ) 0(11)

M e(Hig High None (IGSCC) Low (Medium) B- 2 01 0 JPI 4 (2)

(High) B-J 2 0 0 MS 4 Medium High None Low B-J 105 11 11 MS 6 Low Medium None Low B-J 8 0 0 High (High) High CC, (IGSCC) MediuB-F 10 3 5( 4 12 )

RCR 2 (2)

(Medium)

Medium B-F 2 1(5) 1(5)

RCR 4(2) (High High None (IGSCC) Low (Medium) B- 71 76 8(6 11, 2 (High) B-J 71 RCR 4 Medium High None Low B-J 38 4 4 Medium B-F 3 1(7) 1(7)

RHR 4 (2) (High High None (IGSCC) Low (Medium) B- 3 0 0 (High) B-J 3 0 0 RHR 4 Medium High None Low B-J 22 3 3 RHR 6 Low Medium None Low B-J 46 0 0

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Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Consequence Failure Potential Code Weld Original Revised System Category Rank Rank DMs Rank Category Count RI-ISI Other(l' RI-ISI Other(l Medium 1(8) 1(8)

CS 2 (2) High (High)

Medium High CC, (IGSCC) (Medium M(9) B-F 2 1) 1((9) 1)

CS 4 (2) Mdium High None (IGSCC) Low (Medium) B-F 2 1)

(High) I I I_I_I CS 4 Medium High None Low B-J 24 3 3 CS 6 Low Medium None Low B-J 18 0 0 Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure Potential Weld Original Revised d Count Sym Consequence Category t System Category Rank Rank Rank ank RSI Other RI-ISI Other'1' RI-ISI Other HPCI 4 Medium High None Low B-J 12 2 2 HPCI 6 Low Medium None Low B-J 2 0 0 RCIC 4 Medium High None Low B-J 14 2 2 RCIC 6 Low Medium None Low B-J 2 0 0

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Table 1 Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure Potential Weld Original Revised d Count Consequence Category System Category Rank Rank Rank R an k DMs DMs Rank Rank RI-ISI (2)

-SI Other RI-ISI Other(1 RI-ISI Other(1)

Medium B-F B- 2 6 0

10) 0 1()

RWCU 4 (2) Medim High None (IGSCC) Low (Medium)

RWCU 4 Medium High None Low B-J 53 6 6 RWCU 6 Low Medium None Low B-J 11 0 0 FW 2 High High TASCS, CC Medium B-J 12 6 6 FW 2 High High TASCS Medium B-J 23 3 3 FW 4 Medium High None Low B-J 82 9 9 FW 6 Low Medium None Low B-J 6 0 0 Notes are shown in parenthesis:

1. The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-112657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, Fermi 2 achieved greater than a 10% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

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Notes for Table 1 (cont'd):

2. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
3. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
4. These three welds were selected for examination by both the IGSCC Program and the RI-ISI Program. Since crevice corrosion was identified along with IGSCC as a potential damage mechanism for these welds, the examinations will include the requirements identified in EPRI TR-112657 for crevice corrosion examinations in order to be credited toward both the IGSCC and the RI-ISI Programs.
5. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
6. These welds were selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for these welds, the IGSCC examinations will be credited toward both programs.
7. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.
8. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since crevice corrosion was identified along with IGSCC as a potential damage mechanism for this weld, the examination will include the requirements identified in EPRI TR-112657 for crevice corrosion examinations in order to be credited toward both the IGSCC and the RI-ISI Programs.

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9. This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.

10.This one weld was selected for examination by both the IGSCC Program and the RI-ISI Program. Since IGSCC was the only potential damage mechanism identified for this weld, the IGSCC examination will be credited toward both programs.

11 .A weld previously selected was determined to have limitations. An alternate weld was selected.

12.Additional single side access welds were selected to compensate for dual side welds that could not be examined due to configuration.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 14 Third Interval Relief Request (Continued) 1 Document basis for exclusion of system alignments IE-A4a Not Met None. This gap is a result of the need to (human induced failures). Refer to consensus document the exclusion of human induced report EPRI 1013492 Table 5-3; Human induced failures. It is not anticipated that modeling Initiators are included in the generic priors used changes will be necessary to address this from NUREG/CR-5750. SR.

2 Perform a review of plant data (e.g. CARDs, Daily IE-A7 Met (I) Not Significant. Precursor review is Work request) for the purpose of identifying unlikely to significantly alter initiators additional initiating events, relevant to RI-ISI. This SR was ranked as a CC I. This is in keeping with the ranking required for RI-ISI.

3 Update CAFTA model with updated IE frequencies IE-Cla Not Met Not significant. A review of the updated IE that were generated using the Bayes methodology, frequencies revealed little change to those initiators relevant to RI-ISI. The consequence rank for the initiators will not change from that in the FermiV7 model.

4 SR IE-C10 directs to compare and explain IE-C10 Not Met Not Significant. Current IE frequency differences in IE frequency results with generic development is consistent with industry sources. Such documentation is not provided in the standards. A cross comparison with other EF2 IE Notebook. Provide a comparison of EF2 data sources is primarily a concern for frequencies to industry generic sources in the documentation only.

Summary section of the EF2 IE Notebook (or in an appendix). One can use NUREG/CR-5750 values.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 15 Third Interval Relief Request (Continued) 5 Develop a final report (or database) which can be IE-D1 Not Met None. This is a documentation issue. It is referenced to demonstrate that each of the ASME not anticipated that the model will be elements has been considered. modified to address this item.

6 Complete the "Uncertainty Analysis" section IE-D3 Not Met Not Applicable. The EPRI RI-ISI process is associated with each section of the PRA analysis, structured such that model uncertainties will The EF2 PSA model-of-record and associated not unduly influence results. Also, Draft documentation need to be enhanced to include EPRI Report, "PRA Technical Adequacy sensitivity assessments of key assumptions and Guidance for Risk-Informed Inservice parameters (e.g., the list of sensitivities Inspection Programs." states that this incorporated into the BWROG Option 2 Guidelines requirement need not be met for RI-ISI are generically applicable). applications.

7 Develop a final report (or database) which can be AS-C1 Not Met None. See Gap #5.

referenced to demonstrate that each of the ASME elements has been considered.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 16 Third Interval Relief Request (Continued) 8 Documentation requires updating and corrections to AS-C2 Not Met None. See Gap #5.

be better aligned with latest model. It contains outdated information. For example (but not limited to):

1. Suppression pool must be initiated by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> not 6 as stated in the document.
2. Documentation refers to several Event Trees (ET) where the current model has only one ET.
3. Discussion in 4.2.3 regarding RISKMAN model may only lead to confusion. It may be better to remove references to RISKMAN Event Trees.
4. Documentation indicates a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> short-term HPCI/RCIC success, but the current model utilizes a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> mission time.

9 Add Accident Sequence Analysis section to the AS-C3 Not Met Not Applicable. See Gap #6.

"Uncertainty Analysis" section associated with the each section of the PRA analysis.

10 Update mission time for exceptions to the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SC-A5 Met (I) None. This SR was ranked as a CC I. This limit to be internally consistent (e.g., HPCI/RCIC 4 is in keeping with the ranking required for versus 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />). No other actions are required. RI-ISI in Draft EPRI Report, "PRA Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."

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PRA S(amlard to RI-ISI Sulmfttal Ap,licAbflitv of Weitified Gips to Capability CategorN (CC) ll of ASMI1 11 Some reasonableness checks have been performed SC-B5 Not Met Not Significant. Due to the conservatism (MSPI system success Criteria calculation results inherent in the EPRI RI-ISI methodology, are comparable to MAAP calculation results), the completion of reasonableness checks on Complete reasonableness check and acceptability of all evaluations is judged to be non-the results of the thermal/hydraulic, structural, or significant. It should be noted that there other supporting engineering bases used to support were no significant deviations identified for the success criteria. the calculations where reasonableness checks have already been performed.

12 Develop a final report (or database) which can be SC-C1 Not Met None. See Gap #5.

referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models.

Follow their lead and prepare a Fermi specific document Consider creating success criteria sections similar to MSPI systems for all systems/ functions modeled in the PRA.

13 Complete the "Uncertainty Analysis" section SC-C3 Not Met Not Applicable. See Gap #6.

associated with the each section of the PRA analysis.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 18 Third Interval Relief Request (Continued) 14 Review in detail with system engineers and SY-A4 Not Met Not Significant. It is not anticipated that operators to confirm that the systems analysis this process will have a significant impact correctly reflects the as-built, as-operated plant. on the quantification results with respect to Document this review. RI-ISI.

15 The PRA model is judged to include proper SY-A12 Not Met Not Significant. Due to the conservatism treatment of components and failure modes for inherent in the EPRI RI-ISI methodology, Capability Category II requirements. Additional the lack of documentation of a search for investigation in determining whether all appropriate additional equipment and failure modes is components and failure modes are included could judged to be non-significant.

be performed; however, this is judged not to have significant beneficial impact on the model.

To fully address this SR it is necessary to document a search for additional component and failure modes.

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Attachnent I 16 The PRA model is judged to include proper SY-A14 Not Met Not Significant. It is not anticipated that treatment of components and failure modes for this process will have a significant impact Capability Category II requirements. Additional on the quantification results with respect to investigation in determining whether all appropriate RI-ISI.

components and failure modes are included could be performed; however, this is judged not to have significant beneficial impact on the model.

In order to explicitly meet this SR it would be necessary to provide documentation of why components or failure modes are not included in the model utilizing the screening criteria cited by this SR.

17 The current HVAC dependency analysis uses SY-B 13 Not Met Not Significant. The level of detail subjective judgment or operator action as a basis for provided in the subjective HVAC excluding modeling HVAC for certain areas. This dependency analysis cited is adequate to was appropriate for the IPE analysis but this address the RI-ISI application.

analysis should be improved to meet the ASME standard.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 20 Third Interval Relief Request (Continued) 18 Develop a final report (or database) which can be SY-C1 Not Met None. See Gap #5.

referenced to demonstrate that each of the ASME elements has been considered.

There is no documented independent review of system notebooks. Develop system notebooks and have them reviewed.

19 Many system notebooks do not address several SY-C2 Not Met None. See Gap #5.

ASME SRs (i.e., System Fault trees, evidence of system engineer review etc.).

Improve documentation by explicitly addressing as many of the supporting requirements as possible.

Document interviews with system engineers and plant operators that confirm that the system analysis reflects the as-built, as-operated plant.

Update dependency matrix utilizing CAFTA nomenclature. Also revise to include HVAC support and other systems included in the PRA model.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 21 Third Interval Relief Request (Continued)

A,ttachinent 20 Complete the "Uncertainty Analysis" section SY-C3 Not Met Not Applicable. See Gap #6.

associated with each section of the PRA analysis.

21 Update the review of test and maintenance HR-B1 Not Met Not significant. Pre-initiator human actions procedures. Document the additional rules utilized are judged to not significantly impact RI-ISI for screening individual activities from further applications.

consideration.

22 Document review of activities that may cause HR-C2 Met (I) None. This SR was ranked as a CC I. This failure of automatic realignment (e.g., EDG start is in keeping with the ranking required for signals). Include a review of other failure modes RI-ISI in Draft EPRI Report, "PRA identified during update of the unavailability Technical Adequacy Guidance for Risk-analysis Informed Inservice Inspection Programs."

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 22 Third Interval Relief Request (Continued) 23 A review of HRA documentation has uncovered HR-E3 Not Met Not Significant. The Top 30 risk significant some discrepancies regarding the referenced operator actions (by RAW) have been procedures (CARD 07-21694). Talk through (i.e. reviewed by the operating and training staff.

review in detail) with plant operators to confirm The remainder of the human actions are of that the HR analysis correctly reflects the as- low significance.

operated plant. Document this review.

Establish a method to monitor procedure changes. Not Applicable. This gap relates to capturing a method to ensure procedures are updated with respect to their impact on the HRA. This is a program maintenance and documentation issue which does not affect the current RI-ISI submittal.

24 Significant operator actions should be reviewed in HR-E4 Met (I) Not Significant. The Top 30 risk significant detail by the operating staff and their impact operator actions (by RAW) have been included in the HRA evaluation. Use simulator reviewed by the operating staff. The observations or talk-through with operators to remainder of the human actions are of low confirm the response for scenarios modeled. significance.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 23 Third Interval Relief Request (Continued) 25 Significant operator actions should be reviewed in HR-G5 Met (I) None. This SR was ranked as a CC I. This detail by the operating staff and the associated is in keeping with the ranking required for comments included in the HRA evaluation. Use of RI-ISI in Draft EPRI Report, "PRA simulator observations or talk-through with Technical Adequacy Guidance for Risk-operators to confirm the action times for scenarios Informed Inservice Inspection Programs."

modeled.

26 Develop a final report (or database) which can be HR-I1 Not Met None. See Gap #5.

referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models.

Follow their lead and prepare a Fermi specific document.

27 Document the processes used to identify, HR-I2 Not Met None. See Gap #5.

characterize and quantify the pre-initiator, post-initiator and recovery actions considered in the PRA (including the inputs, methods, and results).

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 24 Third Interval Relief Request (Continued) 28 Complete the "Uncertainty Analysis" section HR-I3 Not Met Not Applicable. See Gap #6.

associated with each section of the PRA analysis.

The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).

29 There is no discussion provided on the type of DA-A2 Not Met Not significant. This is primarily a distribution utilized. Add discussion on this issue documentation issue. The distributions to the PSA documentation, utilized at Fermi are generally consistent with best practices for the failure modes modeled.

30 In order to meet Category II, parameter estimation DA-B1 Met (I) None. This SR was ranked as a CC I. This must group components based on service condition is in keeping with the ranking required for (e.g., clean vs. untreated water, air). RI-ISI in Draft EPRI Report, "PRA Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."

31 Provide evidence that this "interview" process has DA-C12 Met (I) None. This SR was ranked as a CC I. This and continues to occur, and/or perform interviews is in keeping with the ranking required for with plant maintenance and operation staff to RI-ISI in Draft EPRI Report, "PRA ensure a reliable estimate of unavailability for Technical Adequacy Guidance for Risk-significant basic events. Informed Inservice Inspection Programs."

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 25 Third Interval Relief Request (Continued)

Ga .c...ri.tono o f (a) SR ,p cc .nc. 1, or. to R 1-IST 32 The approach suggested is to identify dominant risk DA-C13 Not Met Not significant. The model is reasonably contributor combinations based on knowledge of consistent with known plant operating the accident sequence modeling, and model such practice and experience. An exhaustive combinations of coincident maintenance outages in assessment is not needed to support the use the fault tree logic. A review of recent maintenance of PRA for RI-ISI.

experience should then be performed to identify events of coincident maintenance outages for these combinations to support probability estimation for the events 33 There is no evidence that this check was made. DA-D4 Met (I) None. This SR was ranked as a CC I. This Check that the posterior distribution is reasonable is in keeping with the ranking required for given the relative weight of evidence provided by RI-ISI in Draft EPRI Report, "PRA the prior and the plant-specific data. Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."

34 Data analysis section when compared to other DA-E1 Not Met None. See Gap #5.

elements is the most complete but it is still in draft form. Develop a final report (or database) which can be referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models. Follow their lead and prepare a Fermi specific document.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 26 Third Interval Relief Request (Continued) 35 Complete the "Uncertainty Analysis" section DA-E3 Not Met Not Applicable. See Gap #6.

associated with each section of the PRA analysis.

The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).

36 Document the screening process and results to IF-Bib Not Met Not Applicable. The internal flooding PRA reduce the scenarios which require further analysis. model was not used to support the development of the Fermi RI-ISI Program for Class I piping.

37 To meet strict reading of SR IF-B3, the IF-B3 Not Met Not Applicable. See Gap #36 characteristics (e.g., pressure and temperature, etc.)

of the various flooding sources and mechanisms need to be documented.

38 PERFORM necessary engineering calculations for IF-C3c Not Met Not Applicable. See Gap #36 flood rate, time to reach susceptible equipment, and the structural capacity of SSCs in accordance with the applicable requirements described in Table 4.5.3-2(b) of the ASME standard.

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Ajflicibility (41dewified Gaps to Ca ifitN Catc,ory (CC) 11 of AS11I PWX Stamljird to RI-ISt Sulmittal 39 Update IE frequency utilizing the applicable IF-D5 Not Met Not Applicable. See Gap #36 requirements in Table 4.5.1-2 of the ASME standard. Update flood initiator frequency utilizing updated generic sources, also include uncertainty values.

40 Perform and document the scenario screening IF-D7 Not Met Not Applicable. See Gap #36 process.

41 Screen out a flood area if the product of the sum of IF-E3a Not Met Not Applicable. See Gap #36 the frequencies of the flood scenarios for the area, and the bounding conditional core damage probability (CCDP) is less than 10-9/reactor year.

The bounding CCDP is the highest of the CCDP values for the flood scenarios in an area.

42 The internal flooding HEP calculations need to be IF-E5 Not Met Not Applicable. See Gap #36 enhanced to be consistent with the requirements of HLR HR. These HEP calculations should be integrated into the HRA notebook and performed using the EPRI HRA Calculator.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 28 Third Interval Relief Request (Continued) 43 There are no HRA Calculator entries related to IF-E5a Not Met Not Applicable. See Gap #36 flood scenarios. Conservative estimates are used without details on how they were estimated (e.g.

A3G10-SM-2).

Two types of HRAs should be evaluated.

1. Review HRA values that are in the CAFTA model that would be impacted by the flood scenarios. If necessary copy and modify values to include flood initiator impact.
2. Enter necessary data into HRA Calculator for flood specific HRAs. These HEP calculations should be integrated into the HRA Notebook.

44 Only six flood initiators are included in the Fermi IF-E6 Not Met Not Applicable. See Gap #36 fault tree model. Other initiators should be added if they do not meet the flood screening criteria. The number that should be added is dependent on completing the flood screening process.

45 Perform review of LERF calculations for IF-E7 Not Met Not Applicable. See Gap #36 unscreened flood scenarios.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 29 Third Interval Relief Request (Continued)

Attacimeit I 46 Develop a final report (or database) which can be IF-F1 Not Met None. See Gap #5.

referenced to demonstrate that each of the ASME elements has been considered. Exelon has prepared a matrix type document for all their PRA models.

Follow their lead and prepare a Fermi specific document. Update documentation to reflect CAFTA model nomenclature.

47 Complete the "Uncertainty Analysis" section IF-F3 Not Met Not Applicable. See Gap #6.

associated with each section of the PRA analysis.

The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 30 Third Interval Relief Request (Continued) 48 Document the review of a sampling of non- QU-D4 Not Met Not Significant. Due to the conservatism significant accident cutsets to ensure they are inherent in the EPRI RI-ISI methodology, reasonable and have physical meaning. the lack of documentation of a review of non-significant accident sequences is judged to be negligible. It should be noted that an aggregate review of all accident sequences was performed prior to release of the current model.

49 Provide documentation that includes identification QU-D5a Met (I) None. This SR was ranked as a CC I. This of significant common cause failure contributors to is in keeping with the ranking required for CDF. To attain CC II, initiator fault trees would RI-ISI in Draft EPRI Report, "PRA have to be developed quantified to present Technical Adequacy Guidance for Risk-importance values. Informed Inservice Inspection Programs."

50 Complete the "Uncertainty Analysis" section QU-El Not Met Not Applicable. See Gap #6.

associated with each section of the PRA analysis.

The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 31 Third Interval Relief Request (Continued)

GLII - Des,cripoio of Gap SR I,, 1m$ortanc to RMSt 51 Complete the "Uncertainty Analysis" section QU-E2 Not Met Not Applicable. See Gap #6.

associated with each section of the PRA analysis.

The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).

52 To be consistent with SR QU-E4, the EF2 PRA QU-E4 Met (I) None. This SR was ranked as a CC I. This model and associated documentation should be is in keeping with the ranking required for enhanced. The enhancement should include RI-ISI in Draft EPRI Report, "PRA sensitivity assessments of the impact of uncertain Technical Adequacy Guidance for Risk-model boundary conditions and logical Informed Inservice Inspection Programs."

combinations of uncertainties.

The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 32 Third Interval Relief Request (Continued) plicabilitN _11s t " ability Cae!gor C)11 of ASI11PRA S(anaad to RI-ISI Subtit 53 Develop a final report (or database) which can be QU-F1 Not Met None. See Gap #5.

referenced to demonstrate that each of the ASME elements has been considered.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 33 Third Interval Relief Request (Continued) 54 Strict reading of SR QU-F1 would indicate that the QU-F2 Not Met None. See Gap #5.

following enhancements to the documentation of the EF2 PRA would need to be made to comply with the Standard:

(c) Add a general description of the quantification process including accounting for systems successes.

(d) Document the process and results for establishing the truncation screening values for final quantification demonstrating that convergence towards a stable result was achieved.

(g) Add a discussion on equipment or human actions that are the key factors in causing the accidents to be non-dominant.

(h) Perform and document all sensitivity studies.

(i) Update the uncertainty distribution for the total CDF (3-33)

(1) Document asymmetries in quantitative modeling to provide application users the necessary understanding regarding why such asymmetries are present in the model.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 34 Third Interval Relief Request (Continued) 55 Complete the "Uncertainty Analysis" section QU-F4 Not Met Not Applicable. See Gap #6.

associated with each section of the PRA analysis.

The EF2 PSA model-of-record and associated documentation need to be enhanced to include sensitivity assessments of key assumptions and parameters (e.g., the list of sensitivities incorporated into the BWROG Option 2 Guidelines are generically applicable).

56 Document limitations in the quantification process QU-F5 Not Met None. See Gap #5.

that would impact applications 57 DEVELOP system models that support the accident LE-C5 Not Met Not Significant. This gap is a result of progression analysis consistent with the applicable some of the Level 2 support systems not requirements for paragraph 4.5.4, as appropriate for being modeled and/or documented to the the level of detail of the analysis. same standards as the Level 1 model. It is judged that the level of system modeling or alternate approximations utilized at Fermi are adequate to support the RI-ISI application.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 35 Third Interval Relief Request (Continued) 58 Some of the Level 2 HRA analysis is not performed LE-C6 Not Met Not significant. This is primarily a to the same level of detail as Level 1 HRAs. documentation quality issue for the Level 2 HRA basic events. The EPRI methodology Add the same level of details to the Level 2 HRAs bounds many of the uncertainties inherent in as included in Level 1 HRAs. Ensure updated the Level 2 modeling.

procedures are utilized for this HRA analysis.

Adjust HRA values as appropriate.

59 Review significant LERF sequences and evaluate if LE-C9b Met (I) None. This SR was ranked as a CC I. This an engineering analysis can support equipment is in keeping with the ranking required for continued operation or operator action after RI-ISI in Draft EPRI Report, "PRA containment failure. Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."

60 Review significant LERF sequences and evaluate if LE-C10 Met (I) None. This SR was ranked as a CC I. This an engineering analysis can support equipment is in keeping with the ranking required for continued operation or operator action after RI-ISI in Draft EPRI Report, "PRA containment failure. Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."

61 PERFORM a realistic interfacing system failure LE-D3 Met (I) None. This SR was ranked as a CC I. This probability analysis for the significant accident is in keeping with the ranking required for progression sequences resulting in a large early RI-ISI in Draft EPRI Report, "PRA release. Technical Adequacy Guidance for Risk-Informed Inservice Inspection Programs."

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 36 Third Interval Relief Request (Continued)

ApliRcahilit' ot'We'ified G-41n to CapabilitY CaygorN (CC) II ol'AS-Nl PZA Stanarid to 10-IS1 SOu nittal 62 Document all Level 2 equipment failures and LE-E1 Not Met None. See Gap #5.

operator actions consistent with Level 1 requirements.

63 Strict reading of SR LE-E4 would indicate that the LE-E4 Not Met Not significant. Do to the conservative following enhancements to the LERF analysis and nature of the RI-ISI process, Level 2 HEP associated documentation would need to be made to dependencies are judged not to be comply with the Standard: significant. The EPRI methodology bounds

  • Explicitly assess dependencies among Level 2 many of the uncertainties inherent in the HEPs (and combinations of Level 2 HEPs with Level 2 modeling.

Level 1 HEPs)

  • Complete quantitative uncertainty assessment of the LERF analysis (refer to Disposition for SR LE-F2). Sensitivity runs should be updated.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 37 Third Interval Relief Request (Continued) 64 Lack of traceability from current CAFTA model to LE-G1 Not Met None. Additional documentation and model Level 2 (IPE) documentation. Level 2 analysis nomenclature changes are needed to close refers to the large event trees used with the this gap. It is not anticipated that the model RISKMAN calculation. Currently the CAFTA nomenclature changes will affect the model is a large fault tree model. Documentation quantification results.

should address this difference in methodology and be revised accordingly.

1. Rename Level 2 Basic Event to be consistent with Level 1 BEs.
2. Construct a translation matrix between the CAFTA Model and the Level 2 (IPE) Report.
3. Create a CAFTA Level 2 Report.

65 A review of Work Request (WR) database which MU-B2 Not Met Not Applicable. The prioritization of open tracks model changes has inconsistent priority (1-4) model work requests in the PSA Work given to significance (A-D) ratings. Tracking Database (WTDB) has no impact Establish a consistent method for assigning priority on the results used to support the RI-ISI based on significance. Ensure significance ofWRs program.

is properly assessed. Re-evaluate WR in the database and assign higher priority to model changes that have potential of having the highest impact on the core damage risk.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A30 Page 38 Third Interval Relief Request (Continued)

Attcadinient I ApL)ficab)ifity ot hicuified Gaps to Captbflit8 CateioiN (CC) 11 of ASall PRA Stanlair] to RI-ISI Sulmittal 66 The EF2 PRA received a PRA peer review in 1997 MU-B4 Not Met None. The Self-Assessment (Gap Analysis) using the BWROG peer review guidelines. The process is being used to define the current PRA was converted from the RISKMAN software Fermi PRA technical adequacy for the RI-platform. A new peer review should be performed ISI submittal. A Peer Review is currently to address both 1) the change in software platform, scheduled for the Spring of 2010.

and 2) the more stringent PRA peer review requirements per the ASME PRA Standard (Addendum B).

67 The PRA maintenance and update procedures MU-C1 Not Met Not Applicable. The lack of explicit should include guidelines for the cumulative impact guidelines for assessing the cumulative of pending changes on risk applications. This can impact of pending changes on risk be done during the WR evaluation process. applications does not impact the RI-ISI application. It should be noted that the RI-ISI results are reviewed after every major model release.

Notes:

1. SR: Supporting Requirement
2. SR Codes are contained in ASME RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," December 2005.

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A36 Page 39 Third Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)

-Hardship or Unusual Difficulty without a Compensating Increase in Level of Quality or Safety-

1. ASME Code Components Affected ASME Code Class: Code Class 1

References:

ASME Section XI, Table IWB-2500-1 and IWB-5222 Examination Category: B-P (All Pressure Retaining Components)

Item Number: B15.10

Description:

Alternative Pressure Testing Requirements for the RPV Flange Leak-Off Piping Components: NPS 1 RPV Flange Seal Leak-Off Piping

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through 2003 Addenda

3. Applicable Code Requirement

IWB-2500, Table IWB-2500-1, Code Category B-P, Item Number B15.10 requires that all Class 1 pressure retaining components be Visual, VT-2 examined each refueling outage. The required system pressure test can be either a hydrostatic test or a system leakage test. The system leakage test is performed at a pressure not less than the pressure corresponding to 100% rated reactor power. Per IWB-5222(a), the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second closed valve at the boundary extremity. Per IWB-5222(b), the pressure retaining boundary during the system leakage test conducted at or near the end of the interval shall extend to all Class 1 pressure retaining components within the system boundary.

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A36 Page 40 Third Interval Relief Request (Continued)

4. Reason for Request

As discussed in 3, "Applicable Code Requirements", ASME Section XI, 2001 Edition through 2003 Addenda requires that Class 1 pressure boundary piping shall be pressure tested after each refueling outage. The Reactor Pressure Vessel (RPV) head flange seal leak detection piping is separated from the reactor coolant pressure boundary by one passive membrane, which is an O-ring located on the inner vessel flange as shown in Attachment 1.

A second O-ring is located on the outside of the tap in the vessel flange. Failure of the inner O-ring is the only condition under which this line is pressurized. Therefore, the line is not expected to be pressurized during the system pressure test following a refueling outage.

The configuration of this piping precludes system pressure testing while the vessel head is removed because the configuration of the vessel tap coupled with the high test pressure prevents the tap in the flange from being temporarily plugged or connected to other piping.

The opening in the flange is smooth walled, making the effectiveness of a temporary seal very limited. Failure of a temporary test seal could possibly cause ejection of the device used for plugging or connecting to the vessel flange. To perform the system leakage test in accordance with the Code requirements, the RPV head flange seal detection piping would have to be redesigned, fabricated, and installed. This would impose a severe and unnecessary burden.

The configuration also precludes pressurizing the line with the head installed because the seal prevents complete filling of the piping, which has no vent available. The top head of the vessel contains two grooves that hold the O-rings. The O-rings are held in place by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test were to be performed with the head on, the inner O-ring would be pressurized in a direction opposite to its design function. This test pressure would result in a net inward force on the inner O-ring that would tend to push it into the recessed cavity that houses the retainer clips.

The thin O-ring material would very likely be damaged by the inward force. The design of this line makes the ASME Code required system leakage test impractical either with the vessel head installed or removed.

5. Proposed Alternative and Basis for Use In lieu of the requirements of IWB-5222(b), a VT-2 visual examination will be performed on the subject piping during vessel flood-up each refueling outage. The hydrostatic head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross indications in the piping.

The flange seal leak-offline is essentially a leakage collection and detection system. The line would only function as a Class 1 pressure boundary if the inner O-ring fails thereby pressurizing the line. During this time, the control room annunciator would be in alarm. If the annunciator ceases to be in alarm, this would indicate that the outer O-ring or seal leak-off line had failed and resulted in a reactor coolant pressure boundary leak. This would

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A36 Page 41 Third Interval Relief Request (Continued) require immediate plant shutdown. Per IWA-5243, "when leakage from components are normally expected and collected (such as valve stems, pump seals, or vessel flange gaskets) the visual examination VT-2 shall be conducted by verifying that the leakage collection system is operative."

Fermi 2 has implemented a periodic Preventive Maintenance Event (PM Event B564) to pressurize an isolable section of the leak detection line to verify that the pressure switch is operative and in calibration. Performance of this event satisfies the intent of IWA-5243 for leakage collection systems.

This line is also inspected during the VT-2 system leakage test. There has been no report of gross structural defect to date. Additionally, the line is filled at static head pressure during the outage while the reactor cavity is flooded. Any gross leakage would be easily detected during drywell entries. The VT-2 is considered to be met by PM Event B564 along with a visual inspection of the line during flood-up.

Pursuant to 10 CFR 50.55a(a)(3)(ii), approval is requested to use the proposed alternative described above in lieu of the Code requirement on the basis that the Code requirements establish a hardship or unusual difficulty without a compensating increase in the level of quality or safety.

6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019.
7. Precedents Fermi 2 Relief Request RR-A36, "Evaluation of Second 10-Year Interval Inservice Inspection Request for Relief No. RR-A36 on end of Interval System Pressure Test" (TAC No. ME0868) as approved by the NRC in a letter dated August 28, 2009.

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A36 Page 42 Third Interval Relief Request Attachment 1 Reactor Pressure Vessel Seal Leak-Off Details Detroit Edison Fermi 2

Enclosure to 10 CFR 50.55 a Request NRC-10-0004 Number RR-A36 Page 43 Third Interval Relief Request Attachment 1 (Continued)

N See Figure 2 for detail REATORPRESUR VESELHEAD FLANGE LEAK-OFF LINE CONFIGURATION

Enclosure to 10 CFR 50.55a Request NRC-10-0004 Number RR-A36 Page 44 Third Interval Relief Request Attachment 1 (Continued)

Outer Flange High Pressure Leak Detection Monitoring Tap Seal Ring-, L

- Inner Flange Seal Ring 4 See Detail "A" Detail W*"

Vessel Flange Sectional View Figure 2 REACTOR PRESSURE VESSEL HEAD FLANGE LEAK-OFF LINE DETAILS

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A37 Page 45 Third Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Components Affected ASME Code Class: Code Class 1

References:

ASME Section XI, 2001 Edition with 2003 Addenda Code Case N-702 BWRVIP-108: BWR Vessel and Internals Project, "Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1003557, October 2002 Examination Category: B-D Item Numbers: B3.90, B3.100

Description:

Alternative Requirements for Examination of Boiling Water Reactor (BWR) Nozzle Inner Radius Sections and Nozzle-to-Shell Welds Components: N1, N3, N5, N6, N7, N8, and N10 Nozzles (see Attachment 2 for specific nozzle identifications)

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through 2003 Addenda

3. Applicable Code Requirement

Table IWB-2500-1, Examination Category B-D, Inspection Program B, Item Numbers B3.90, and B3.100 require a volumetric examination to be performed once per interval on each reactor vessel nozzle-to-vessel weld and nozzle inside radius section.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A37 Page 46 Third Interval Relief Request (Continued)

4. Reason for Request

Leverage the technical basis and criteria in BWRVIP-108 to realize substantial radiation dose and cost savings while maintaining an acceptable level of quality and safety.

5. Proposed Alternative and Basis for Relief Pursuant to 10CFR50.55a(a)(3)(i), Detroit Edison requests authorization to utilize the alternative requirements described herein in lieu of the requirements of Table IWB-2500-1, Examination Category B-D, Item Numbers B3.90, and B3.100 for reactor vessel nozzle-to-shell welds and nozzle inside radius sections. These alternative requirements will not be utilized for the feedwater nozzles or the recirculation inlet nozzles. The alternative requirements will be applied to the control rod drive return nozzle which has been cut and capped at Fermi 2 and therefore is not subject to the thermal fatigue issues which might otherwise be a concern for operational control rod drive return lines.

This alternative allows a 25% sampling of the reactor vessel nozzle inner radius section examinations and nozzle-to-shell weld examinations to be implemented, provided at least one nozzle from each system and nominal pipe size is examined. This alternative also allows a VT-1 examination of the nozzle inner radius base metal surfaces to be performed in lieu of the Code required volumetric examination. For the nozzle-to-shell welds requiring examination, a volumetric examination will be performed. ASME Section XI, Appendix VIII, 2001 Edition with no Addenda will be used for volumetric examinations as mandated in 10CFR50.55a(b)(2)(xv).

Current requirements for the inspection of the BWR nozzle-to-shell welds and nozzle blend radii are costly and result in significant radiation exposure to examiners. The performance of NDE has improved substantially since the examinations of ASME Section XI, Table IWB-2500-1, Examination Category B-D, Item Numbers B3.90, and B3.100 were first required such that there is now a high reliability of detecting flaws that can challenge the structural integrity of BWR nozzles and their associated welds. Knowledge of improved NDE capabilities, coupled with fracture mechanics, provides a technical basis to justify reduction of inspections while maintaining safety. This technical basis is provided in BWR Vessel and Internals Project (BWRVIP) Report No. BWRVIP-108. For any cracks in the nozzle blend radius region, the results of BWRVIP-108 show that the conditional failure probability of the nozzles due to a low temperature overpressure (LTOP) event are very low (<1x10-6 for 40 years), even without any inservice inspection. At the nozzle-to-vessel shell weld, the conditional probability of failure due to the LTOP event is also very small (<1x10-6 for 40 years), with or without any inservice inspection. As such, the BWRVIP-108 report provides the technical basis for the reduction of the nozzle-to-shell welds and nozzle blend radii from 100% to 25% of the nozzles every 10 years.

This EPRI report received an NRC SER dated December 19, 2007. In the SER, Section 5.0 "Plant Specific Applicability" indicates that each licensee who plans to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A37 Page 47 Third Interval Relief Request (Continued) inner radius sections may reference BWRVIP-108 report as the technical basis for the use of the alternatives presented herein. However, each licensee should demonstrate the plant-specific applicability for the BWRVIP-108 report to its units in the relief request by showing that all the general and nozzle-specific criteria addressed below are satisfied (reference Attachment 1):

(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 115 0 F per hour.

a. Per TS SR 3.4.10.1, the RPV heatup/cooldown is limited to less than or equal to 100 0F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. This criterion is met.

(2) For the Recirculation Inlet nozzles, the following criteria must be met:

a. (pr/t)/CRpv<l.15, the calculation for the Fermi 2 N2 Nozzle results in 0.89 which is less than 1.15.
b. [p(ro2+ri2)/(ro2-ri2)]/CNOzzLE< 1.15, the calculation for the Fermi 2 N2 Nozzles results in 1.229 which is greater than 1.15, therefore this criteria is not met for the Recirculation Inlet Nozzles.

(3) For the Recirculation Outlet nozzles, the following criteria must be met:

a. (pr/t)/CRpv<l.15, the calculation for the Fermi 2 N2 Nozzles results in 1.07 which is less than 1.15.
b. [p(ro2+ri2)/(r0o2ri2)]/CNOzzLE <1.15, the calculation for the Fermi 2 N2 Nozzles results in 1.069 which is less than 1.15.

Based upon the above information, all RPV nozzle-to-vessel shell welds and nozzle inner radius sections, with the exception of the Recirculation Inlet Nozzles, meet the criteria and therefore are applicable. Since the Recirculation Inlet Nozzles do not meet the criteria, the alternative requirements will not be applied to those nozzles. Additionally, as stated earlier, this alternative will not be applied to the feedwater nozzles.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A37 Page 48 Third Interval Relief Request (Continued)

6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019.
7. Precedents Duane Arnold Energy Center - Safety Evaluation for Request for Alternative to Reactor Pressure Vessel Nozzle to Vessel Weld and Inner Radius Examinations was approved on August 29, 2008 (Accession number ML080710428, TAC No. MD8193).

Enclosure to 10 CFR 50.55a Request NRC-10-0079 Number RR-A37 Page 49 Third Interval Relief Request (Continued)

Attachment 1 Response to NRC Plant Specific Applicability to BWRVIP-108

1. The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 11i5T/hour Response: TS SR 3.4. 10.1 limits the RPV heatup/cooldown to *1ýl00'F/hour for Curve B and
  • !ý20 0 F/hour for Curve A.

Recirculation Inlet Nozzles Recirculation Outlet Nozzles 2 (pr/t)/CR~pv<1 .15 4 (pr/t)/CRPV<1. .15 p=RPV Normal Operating 1045 p=RLPV Normal Operating 1045 Pressure Pressure r--RPV inner radius 127.125 r-=RPV inner radius 127.125 t=RPV wall thickness 7.6875 t=~RPV wall thickness 7.6875 CRPV= 19332 CRPV= 16171 (Pr/t)/CRPV= 0.89 (pr/t)/CRPV= 1.07 3 [p(r,,2+ri 2 )/(r,,2 -ri 2)]/CNOZZLE<1 .15 5 [P(rol+r, 2)/(ro2 _r, 2)]/CNOZZLE<. .15 p=RPV Normal Operating 1045 p=RPV Normal Operating 1045 Pressure Pressure r,==nozzle outer radius 11I ro=nozzle outer radius 22.5625 ri=nozzle inner radius 6.1875 rj==nozzle inner radius 13.125 CNOZZLE 1637 CNOZ'ZLE 1977

[p(ro +ri2 )/(r. 2-ri 2)]/CNOZZLE= 1.229 [p(r. 2 +ri 2)/(r. 2 _ri 2 )]/CNOZZLE= 1.069

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A37 Page 50 Third Interval Relief Request (Continued)

Attachment 2 Applicable Components Code Item Nozzle Category Number Identification Description System Isometric N3 B-D B3.100 8-316A IRS Nozzle Inside Radius Section Main Steam 5361-5 N3 B-D B3.100 8-316B IRS Nozzle Inside Radius Section Main Steam 5361-5 N3 B-D B3.100 8-316C IRS Nozzle Inside Radius Section Main Steam 5361-5 N3 B-D B3.100 8-316D IRS Nozzle Inside Radius Section Main Steam 5361-5 N5 B-D B3.100 14-316A IRS Nozzle Inside Radius Section Core Spray 5361-5 N5 B-D B3.100 14-316B IRS Nozzle Inside Radius Section Core Spray 5361-5 N10 B-D B3.100 15-315 IRS Nozzle Inside Radius Section Control Rod Drive 5361-5 Reactor Recirc N1 B-D B3.100 5-314A IRS Nozzle Inside Radius Section Suction 5361-5 Reactor Recirc N1 B-D B3.100 5-314B IRS Nozzle Inside Radius Section Suction 5361-5 Jet Pump N8 B-D B3.100 19-314A IRS Nozzle Inside Radius Section Instrument 5361-5 Jet Pump N8 B-D B3.100 19-314B IRS Nozzle Inside Radius Section Instrument 5361-5 N7 B-D B3.100 2-318 IRS Nozzle Inside Radius Section RPV Head Vent 5361-5 N6 B-D B3.100 4-318A IRS Nozzle Inside Radius Section RPV Head Spare 5361-5 N6 B-D B3.100 4-318B IRS Nozzle Inside Radius Section RPV Head Spare 5361-5 N3 B-D B3.90 8-316A M.S. Nozzle-to-Vessel Weld Main Steam 5361-5 N3 B-D B3.90 8-316B M.S. Nozzle-to-Vessel Weld Main Steam 5361-5 N3 B-D B3.90 8-316C M.S. Nozzle-to-Vessel Weld Main Steam 5361-5 N3 B-D B3.90 8-316D M.S. Nozzle-to-Vessel Weld Main Steam 5361-5 N5 B-D B3.90 14-316A C.S. Nozzle-to-Vessel Weld Core Spray 5361-5 N5 B-D B3.90 14-316B C.S. Nozzle-to-Vessel Weld Core Spray 5361-5 CRD Ref. Nozzle-to-Vessel N10 B-D B3.90 15-315 Weld Control Rod Drive 5361-5

Enclosure to 10 CFR 50.55a Request Number RR-A37 NRC-10-0004 Third Interval Relief Request Page 51 (Continued)

Attachment 2 Applicable Components (continued)

Code Item Nozzle Category Number Identification Description System Isometric Reactor Recirc.

N1 B-D B3.90 5-314A RRI Nozzle-to-Vessel Weld Suction 5361-5 Reactor Recirc.

N1 B-D B3.90 5-314B RRI Nozzle-to-Vessel Weld Suction 5361-5 Jet Pump N8 B-D B3.90 19-314A JPI Nozzle-to-Vessel Weld Instrument 5361-5 Jet Pump N8 B-D B3.90 19-314B JPI Nozzle-to-Vessel Weld Instrument 5361-5 Head/Vent Nozzle-to-Vessel N7 B-D B3.90 2-318 Weld RPV Head Vent 5361-5 Spare Nozzle-to-Vessel N6 B-D B3.90 4-318A Weld RPV Head Spare 5361-5 Spare Nozzle-to-Vessel N6 B-D B3.90 4-318B Weld RPV Head Spare 5361-5

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A38 Page 52 Third Interval Relief Request Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Components Affected ASME Code Class: Code Class 1

References:

ASME Section XI, Table IWB-2500-1 ASME Section XI, Appendix VIII, Supplements 4 and 6, 10CFR50.55a(b)(xv).

Examination Category: B-A Item Numbers : B1.30 and B1.40

Description:

Implementation of Appendix VIII, Supplements 4 and 6 - Use of PDI Qualified Procedures, Personnel and Equipment for Non-Appendix VIII RPV Shell-to-Flange Weld and Head-to-Flange Weld.

Components: Reactor Pressure Vessel (RPV) Shell-to-Flange Weld No.13-308 Reactor Pressure Vessel (RPV) Head-to-Flange Weld No. 3-319

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition (No Addenda) per 10CFR50.55a(b)(2)(xv) for Appendix VIII.

3. Applicable Code Requirement

The 2001 Edition with No Addenda of the American Society of Mechanical Engineers (ASME Code)Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Subsection IWA-2232, requires ultrasonic (UT) examinations be performed in accordance with Mandatory Appendix I. Paragraph I-2110(b) of Mandatory Appendix I requires that examination of the RPV shell-to-flange weld and head-to-flange weld to be in accordance with ASME Code,Section V, Article 4.

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A38 Page 53 Third Interval Relief Request (Continued)

4. Reason for Request

The use of this alternative will allow the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel RPV flange welds from the accessible vessel or head side of the welds in accordance with ASME Code,Section XI, Division 1, 2001 Edition, No Addenda, Appendix VIII, Supplements 4 and 6. This alternative would be used in lieu of Article 4 of Section V requirements.

5. Proposed Alternative and Basis for Relief Detroit Edison proposes to perform ultrasonic examinations of the RPV shell-to-flange weld and head-to-flange weld using procedures, personnel, and equipment that have been demonstrated and qualified in accordance with ASME Section XI, 2001 Edition, No Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDI Program. Since the examinations will be performed from a single side due to the weld configuration, all procedures, personnel, and equipment will be qualified for single sided access for examination of these welds.

Appendix VIII requirements were developed and adopted to ensure the effectiveness of ultrasonic examinations within the nuclear industry by means of a rigorous, item specific performance demonstration containing flaws of various sizes, locations, and orientations.

The performance demonstration process has established with a high degree of confidence, the capability of personnel, procedures, and equipment to detect and characterize flaws that could be detrimental to the structural integrity of the RPV. The PDI approach has demonstrated that for detection and characterization of flaws in the RPV the ultrasonic examination techniques are equal to, or surpass the requirements of the ASME Section V, Article 4 ultrasonic examination requirements. Though Appendix VIII is not specified for the RPV flange weld UT examinations, the use of Appendix VIII, Supplements 4 and 6 criteria for detection and sizing of flaws in these welds will be equal to, or exceed the requirements of ASME Section V, Article 4. Therefore, the use of the proposed alternative will continue to provide an acceptable level of quality and safety, and approval is requested pursuant to 10 CFR 50.55a(a)(3)(i).

Enclosure to 10 CFR 50.55a Request Number NRC-10-0004 RR-A38 Page 54 Third Interval Relief Request (Continued)

6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which began on May 2, 2009 and is scheduled to end on May 1, 2019.
7. Precedents A similar relief request (RR ISI-30) was previously approved for Union Electric Company for its Callaway Plant, Unit 1 on April 7, 2004 (ADAMS Accession Nos. ML032340608 and ML041000516).
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