ML20217D996
| ML20217D996 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/13/1999 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20217D994 | List: |
| References | |
| PROC-991013, NUDOCS 9910180229 | |
| Download: ML20217D996 (23) | |
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RISK-INFORMED INSERVICE INSPECTION PROGRAM PLAN - JAMES A FITZPATRICK i l
l Table of Contents
- 2. Proposed Alternative to Current inservice Inspection Programs 2.1 ASME Section XI 2.2 Augmented Programs I
- 3. Risk-Informed ISI Process i
3.1 Scope of Program 3.2 Consequence Evaluation 3.3 Failure Assessment 3.4 Risk Evaluation 3.5 Element and NDE Selection 3.6 Additional Examinations 3.7 Program Relief Requests 3.8 Change in Risk
- 4. Implementation and Monitoring Program
- 5. Proposed ISI Program Plan Change
- 6. References / Documentation 9910180229 gDR ADOCK 0500099101!'
PDR 1
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j INTRODUCTION ,
i 1.1 Relation to NRC Regulatory Guide RG-1.174 !
Inservice inspections (ISI) are currently perfctmed on piping to the requirements of the !
ASME Boiler and Pressure Vessel Code Section XI,1989 Edition as required by l 10CFR50.55a. The unit is currently in the third inspection interval as defined by the l Code for Program B.
The objective of this submittal is to request a change to the ISI program plan for piping through the use of a risk-informed ISI program. The-risk informed process used in this submittal is described in EPRI TP.112657, Final Report, " Risk-Informed Inservice inspection Evaluation Procedurr.."
As a risk-informed application, this submittal meets the intent and principles of !
Regulatory Guide 1.174. Further information is provided in Section 3.7 relative to i
~ defense-in-depth. I 1.2 Individual Plant Examination (IPE) Quality j The Fitzpatrick Level 1 and Level 2 IPE model[ Revision 1, April 1998) was used to I evaluate the consequences of pipe ruptures during operation in Modes 1 and 2.
The base core damage frequency (CDF) and base large, early release frequency (LERF) from this version of the IPE model are 2E-6/yr and 7E-7/yr, respectively.
Revision 1, of the IPE has undergone the BWROG certification process. The results of the certification showed that the IPE "can be effectively used to support Grade 3 applications involving relative risk significance; in addition, absolute risk determination applications can be performed with supporting deterministic analyses." .
l In addition, the NRC reviewed Revision 0 of the IPE, and the following areas for '
improvement were identified:
- 1. Additional candidates for common cause failures.
- 2. Updates to data base to reflect the most recent plant operating experience.
- 3. Estimates used for the likelihood of containment failures at vessel breach due to shell melt-through.
The disposition of these items in the IPE update (Revision 1, April 1998) is discussed in Appendix 1.
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' 2.1 - ASME Section XI I
' ASME Section XI Categories B-F, B-J, C-F-1 and C-F-2 currently contain the I i
requirements for examining (via NDE) piping components. This current program is
!' limited to ASME Class 1 and Class 2 piping. The attemative risk-informed inservice
! inspection (Rl-ISI) program for piping is described in EPRI TR 112657. The RI-ISI l
program will be substituted for the current examination program on piping in accordance with 10 CFR 50.55a(s)(3)(i) by attematively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.
- EPRI TR 112657 provides the requirements defining the relationship between the risk-informed examination program and the remaining unaffected portions of ASME Section i XI. .
2.2 Augmented Programs
' NYPA, together with the BWRVIP and EPRI ere investigating operating experience and !
material performance with respect to the BWR fleet and IGSCC issues. As such, our l response to Generic Letter 88-01 (NUREG-0313, Rev 2) and its supplement remains unchanged, at this time. Two other augmented inspection programs (Generic 89-08; Flow Accelerated Corrosion and Generic Letter 89-13; Service Water) are credited in the RI-ISI program but are not changed by the RI-ISI program. )
l 4 l 3. RISK-INFORMED ISI PROCESSES L The processes used to develop the RI-ISI program are consistent with the methodology
- described in EPRI TR 112657.
The process that is being applied, involves the following steps: i l e Scope Definition e Consequence Evaluation !
'~
+ Failure Assessment :
+ Risk Evaluation
. Element /NDE Selection i e' .lmplement Program I l e. Feedback Loop l
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- There were no significant deviations to the process described in EPRI TR-112657. The only deviation was in the _ element selection process for the Standby Liquid Control (SLC)
. system.
The SLC system had one location in risk c' ategory 4 (high consequence, low failure potential; i.e. no degradation mechanism) and two locations in risk category 5 (medium consequence, medium failure potential; i.e. thermal fatigue). It was decided to inspect both locations in risk category 5 (for thermal fatigue) instead of one location in risk
< category 4 and one location in risk category 5.
The change in risk assessment presented in section 3,8 shows a net reduction in risk with the above taken into consideration.
3.1 Scope of Program -
The system (s) included in the risk-informed ISI program are provided in Table 3.1-1.
The piping and instrumentation diagrams and additional plant information were used to define system boundaries.
3.2 ' Consequence Evaluation The consequence (s) of pressure boundary failures were evaluated and ranked based on their impact on core damage and containment performance (isolation, bypass and large, early release). The impact on these measures due to both direct and indirect effects
, was considered using the guidance provided in EPRI TR-112657.
3.3 - Failure Assessment Failure potential estimates were generated utilizing industry failure history, plant specific failure history and other relevant information. These failure estimates were determined using the guidance provided in EPRI TR 112657. '
I Table 3.3-1 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative.
3.4 Risk Evaluation in the preceding steps, each run of piping within the scope of the program was evaluated to determine its impact on core damage and containment performance (isolation, bypass, and large, early release) as well as its potential for failure. Given the results of these steps,' piping segments are then defined as continuous runs of piping potentially 3'
susceptible to the same type (s) of degradation and whose failure will result in similar consequence (s). Segments are then ranked based upon their risk significance as defined in EPRI TR-112657. -
, The results of these calculations are presented in Table 3.4-1 i 6 '. t -- Page 4 of23 -
- 3.5 Element and NDE Selection l In general, EPRI TR-112657 requires that 25% of the locations in ine high risk regions (i.e. risk categories 1,2 & 3) and 10% of the locations in the medium risk regions (i.e.
risk categories 4 & 5) be selected for inspection and appropriate non-destructive examination (NDE) methods tailored to the applicable degradation mechanism be l'> . defined for ASME Code Case N578 applications. The results of the selection are l ' presented in Table 3.5-1 Section 4 of EPRI TR-112657 was used as guidance in l determining the examination requirements for these locations.
In addition, all in scope piping components, regardless of risk classification, will continue to receive Code required pressure testing, as part of the current ASME Section Xi program. VT-2 visual examinations are scheduled in accordance with the existing pressure test program, which remains unaffected by the risk-informed inservice inspection program.
- l. _ 3.6 Additional Examinations L Since the risk-informed inspection program may require examinations on a number of I
elements constructed to lesser pre-service inspection requirements, the program in all cases will determine through an engineering evaluation the root cause of any unacceptable flaw determined to be service related (i.e., fatigue, wall loss, IGSCC, etc.)
or relevant condition found during examination. The evaluation will include the l applicable service conditions and degradation mechanisms to establish that the element (s) will still perform their intended safety function during subsequent operation. 1 Elements not meeting this requirement will be repaired or replaced. ' ' i The evaluation will include whether other elements on the segment or segments are subject to the same root cause and degradation mechanism. Additional examinations will be performed on these elements up to a number equivalent to the number of ;
l elements required to be inspected on the segment or segments initially. If unacceptable '
flaws determined to be service related or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined. j
, No additional examinations will be performed if there are no additional elements !
identified as being susceptible to the same service related root cause conditions or I degradation mechanism.
3.7 Program Relief Requests l - 1 j Altemate methods are specified to ensum structural integrity in cases where l l examination methods cannot be app!!n due to limitations such as inaccessibility or l 1
radiation exposure hazard.
- A minimum of >90% volume coverage (per Code Case N-460) will be provided, when possible, when performing the risk-informed examinations. However, some limitations will not be known until the examination is performed, since some locations may be ,
= examined for the first time by the specified techniques.
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. At this time, all the risk-informed examination locations that have been selected are estimated to exceed >90% volume coverage. In instances where a location may be found at the time of the examination that does not meet >90% coverage, the process outlined in EPRI TR 112657, Final Report will be followed.
All existing relief requests are unaffected and remain in place. ,
I 3.8 Change in Risk The risk-informed ISI program has been conducted in accordance with Regulatory Guide 1.174, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.
This' evaluation identified the allocation of segments into High, Medium, and Low risk regions of the EPRI TR-112657 and ASME Code Case N578 risk ranking matrix, and then determined for each of these risk classes what inspection changes are proposed for each of the locations in each segment. The changes include changing the number and location of inspections within the segment and in many cases improving the effectiveness of.the inspection to account for the findings of the RI-ISI degradation mechanism assessment. For example, for locations subject to thermal fatigue,
. Inspection locations have an expanded volume and the examination is focused to enhance the probability of detection during the inspection process.
Table 3.8-1 presents a summary of the proposed RI-ISI program versus the current Section XI program. This risk ranking is provided with and without the impact of degradation mechanisms associated with and managed by other augmented inspection programs (e.g.' FAC). The values provided inside parentheses represent the risk category associated with the augmented inspection program. These other augmented programs effectively manage the risk associated with these piping segments unlesc there is the potential for other degradation mechanism (e.g. thermal fatigue) that would not be appropriately managed by these augmented inspections (e.g. FAC). Table 3.8-1 I
identifies on a per system basis each applicable risk category. In addition the following is provided:
. . the consequence rank and degradation mechanism which supports the risk category, e the number of locations inspected by the current section XI, '
- the number of locations proposed for the Rl-ISI program, e the increase, decrease or no change in the number of inspections, e-Page 6 of 23 g
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L ** for risk categories with a decrease between the section XI and RI-ISI programs, an identification of whether those segments are being managed by an augmented inspection program (e.g. IGSCC), and e the risk impact of the RI-ISI program as compared to the section XI program.
l As can be seen from Table 3.8-1, the only high-risk category, as defined in EPRI TR-l 112657, is risk category 2. There are seven systems (ESW, FW, HPCI, MS, RCIC, RHR and RHRSW) with risk category 2 segments. In six of the seven systems there is an increase in the number of inspections required by the RI-ISI program over the current l section XI program. In the seventh system (FW), while the number of inspections is the same,~ more of the risk will be captured as the RI-ISI inspection volumes will be greater than that required by the current section XI program. Thus, a positive impact on risk is expected as each system will see either an increase in the number of inspections or inspection volume.
In the medium risk region (i.e. risk categorica 4 and 5), there are four systems (FW, MS, i
RWCU and RWR) where the number of inspections (RI-ISI vs section XI) has decreased but the system is being managed by an augmented inspection program (FAC or IGSCC). These are identified in Table 3.8-1 by either 'FAC' or 'lGSCC' in the
' Augmented Program' column. As such, the impact of the RI-lSi program for these
- locations is considered negligible as the augmented programs provide any real risk reduction benefit.
, The medium risk region consists of risk categories 4 and 5. Risk category 4 occurs in
! eight systems for the locations not addressed by augmented programs. These are CS, HPCI, MS, RCIC, RWCU, RHR, RHRSW and SLC. Of these, two systems (CS and RHR) have an increased number of inspection (+3), three systems (MS, RCIC and l RWCU) have a decrease in the number of inspections (-4) and three systems (HPCI,
} RHR and SLC) have no change in the number ofinspections. Thus, given a reduction in only one inspection location a negligible impact on risk is expected.
Risk category 5 occurs in eight systems for the locations not addressed by augmented j programs. These are CS, ESW, FW, HPCI, MS, INST, RHR and SLC. Of these, six l systems (ESW, HPCI, MS, INST, RHR and SLC) have an increased number of l' inspections (+9) and two systems (CS and FW) have a decreased number of inspections
- (-6). Thus, given a net gain of three inspections, a risk neutral to positive impact is
' expt.Med.
As discussed in EPRI TR-112657, the contribution to risk from risk category 6 and 7 locations is negligible.
In summary and as depicted in Table 3.8-2, there is an increase of seven inspection ,
locations and four expanded inspection volumes in the high risk region (i.e. risk category 2). There is an increase of two inspection locations in the medium risk region (i.e. risk categories 4 and 5). Thus, there is a risk neutral to positive impact due to the RI-ISI program as compared to the current section XI program, taking into account the existing augmented programs outside of the scope of the section XI program.
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- Qgfense-In-Depth The intent of the inspections mandated by ASME Section XI for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary. Currently, the process for picking inspection locations is based upon structural discontinuity and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev. 1, " Evaluation of Inservice Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds", this method has been ineffective in identifying leaks or failures. EPRI TR-112657 and Code Case N578 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.
This process has two key independent ingredients, that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense in depth is maintained. First off, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased. Secondly, the consequence assessment effort has a single failure criterion.
As such, no matter how unlikely a failure scenario is, it is ranked High in the consequence assessment, and at worst Medium in the risk assessment (i.e., Risk Category 4), if as a result of the failure there is no mitigative equipment available to respond to the event. In addition, the consequence assessment takes into account equipment reliability, and less credit is given to less reliable equipment.
As a result of the above process, the number of inspections has been increased in the high-risk categories in all systems with high-risk segments (only in one system the number ofinspections in high category is unchanged), as can be seen in Table 3.7-1.
The main reduction in number of inspections occurs in low risk categories. All locations within the reactor coolant pressure boundary will continue to receive a system pressure test and visual VT-2 examination as currently required by the Code regardless of its risk classification.
- 4. IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RI-ISI program, procedures that comply with the guidelines described in EPRI TR-112657, Final Report will be prepared to implement and monitor the program. The new program will be integrated into the existing ASME Section XI interval. No changes to the Final Safety Analysis Report are necessary for program implementation.
l The applicable aspects of the Code not affected by this change would be retained, such ;
as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quahty control requirements. Existing ASME Section
- XI program impicmenting procedures would be retained and would be modified to '
address the RI-ISI process, as appropriate. Additionally the procedures will be modified to include the high safety significant locations in the program requirements regardless of their current ASME class.
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l The proposed monitoring and corrective action program will contain the following elements:
-: A. Identify B. Characterize ;
C. (1) Evaluate, determine the cause and extent of the condition identified (2) Evaluate, develop a corrective action plan or plans I
D. Decide
, E. ' implement: {
F. Monitor G. Trend-l The RI-ISI program is a living program requiring feedback of new relevant information to
, ensure the appropriate identification of high safety significant piping locations. As a l minimum, risk ranking of piping segments will be reviewed and adjusted on an ASME :
period basis. In addition, significant changes may require more frequent adjustment as
- directed by NRC Bulletin or Generic Letter requirements, or by industry and plant specific feedback.
i 5.- PROPOSED ISI PROGRAM PLAN CHANGE A comparison between the RI-ISI program and t'he current ASME Section XI program ,
requirements for in-scope piping is given in Table 5-1. An identification of piping l segments that are part of plant augmented programs is also included in Table 5-1. i The initial program will be started in the inspection period current at the time of program approval. For example, the second inspection period of the third inspection interval ends i on September 28,2004. If the program is approved such that a refueling outage l . remains in the second period,66% of the required remaining examinations will be performed by the end of the inspection interval per the risk-informed inspection program.
! 6.0 ' REFERENCES / DOCUMENTATION 6,1 EPRI TR 112657, Final Report, " Revised Risk-Informed inservice Inspection l Evaluation Procedure".
l 6.2 - EPRI TR-106706, Interim Report, " Risk-Informed Inservice inspection Evaluation '
procedure."
6.3 Calculation # NSD-016, Revision 2, Degradation Mechanism Evaluation, dated July 1999.
6.4 Calculation # NSD-017, Revision 1, Consequence Evaluation, dated July 1999. .
6.5 NFGE 99-0030, Risk Ranking and Element Selection Results for the JAF Risk- j informed ISI (RI-ISI) Program, dated August 3,1999, i l
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l l Table 3.1-1 System Selection and Segment Definition System Description Number of Segments !
Control Rod Drive (CRD) 4 Core Spray (CS) 25 Emergency Service Water (ESW) 24 Feedwater(FW) 19 Fuel Pool Cooling (FPC) 1 High Pressure Coolant injection (HPCI) 25 Main Steam (MS) 32 Nuclear Boiler Vessel Instrumentation 6 (INST)
Reactor Core isolation Cooling (RCIC) 15 Reactor Water Cleanup (RWCU) 9 Reactor Water Recirculation (RWR) 60 Residual Heat Removal (RHR) 94 RHR Service Water (RHRSW) 15 Standby Liquid Control (SLC) 4 Total 333 l
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23
Table 3.8-2 l Change In Inspection Summary l
Risk Risk System INSPECTIONS Expanded Region Category Volume Added Deleted Change High 2 ESW l 1
2 MS 2 2 2 RCIC 1 1 2 RHR 1 1 2 RHRSW 1 1 High 2 Total +7 +4 j l
l Medium 4 CS 1 1 4 HPCI 0 4 hiS 1 -1 4 RCIC 2 -2 4 RWCU 1 -1 l 4 RHR 2 2 4 -'
4 Subtotal -1 5 CS 3 -3 I 5 ESW 2 2 !
5- FW 3 -3 ,
5 HPCI 2 2 5 MS 1 1 5 INST 1 1 5 RHR 1 1 5 SLC 2 2
~
5 Subtotal +3 j Medium 4&5 Total +2 i Page 19 of 23
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APPENDIX I J
DISPOSITION OF THE FINDINGS FROM NRC REVIEW OF REVISION 0 OF THE IPE l
i Page 2I of 23
. Finding #1 - additional candidates for common cause failures (Section 2.2),
Response
The update to the James A. Fitzpatrick (JAF) Nuclear Power Plant Individual Plant Examination (IPE) was completed in April 1998. The update incorporated additional common cause failure terms including:
. Common cause equipment failure groups such as fans, check valves, dampers, and transmitters have been included in the analysis.
. Catastrophic common cause failure of both 125V dc battery control boards 71BCB-2A and 71BCB-2B was included as an initiator, which results in a station blackout with loss of HPCI and RCIC and subsequent core damage.
Finding #2 - Updates to the data base to reflect the most recent plant operating experience (Section 2.3.4)
Response
The JAF IPE update consisted of the most recent plant operating data:
e An updated initiating event database, including all scrams that occurred between
//26/1975 and 12/31/1997. l e An updated component failure and unavailability database that reflects failures that l occurred between 1/1/1986 and 4/30/1995 and current on-line maintenance practices.
Finding # 3 - CPI recommendation for estimates used for the likelihood of containment failure at vessel breach due to shell melt-through
Response
The IPE analysis performed to estimate the impact of using drywell sprays (CPI recommendation) to reduce the likelihood of c'rywell(shell) melt-through is as follows:
The operability of drywell sprays during a severe accident can influence both the survivability of the containment and its performance in containing fission products. The IPE contains the following insights:
- 1. The total probability of containment failure decreases because water on the drywell floor reduces the likelihood of drywell liner melt-through and because the sprays reduce containment pressure making static overpressurization less likely.
- 2. Containment failure is delayed. The principal cause for this delaf si the reduction in the likelihood of drywellliner melt-through. This shift will reduce the radiological source term because natural decontamination mechanisms will have more time to act prior to containment failure.
- 3. The location of containment failure shifts slightly from the drywell areas to the wetwell.
The principal cause for this shift is again the reduction in the likelihood of drywell liner melt-through. This shift will reduce the radiological source term because the suppression pool will scrub releases from containment.
I Page 22 of 23
. 4. The sprays will provide direct scrubbing of fission product aerosols and increase L residence time and so enhance the effectiveness of natural decontamination j mechanisms.
[ )
! These insights have been incorporated in the JAF Emergency Operating I Procedures / Severe Accident Operating Guidelines (EOPs/SAOGs). The revised i . EOPs adequately address the use of Drywell So sy. In the SAOGs, drywell
)
l spray operations have been structured for severe accident conditions.
l Specifically:,
l 4
. Drywell spray is prioritized relative to RPV injection (SAOG-1) e Drywell spray initiation is required in the Containment and Radioactivity Release Control Guideline (SAOG-2)
+ ' . Drywell Spray initiation is required for high Drywell radiation. This is useful in ensuring Drywell Spray initiation prior to RPV breach.-
l l
l l
i-L l
Pa6e 23 of 23 i
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