ML103550487
| ML103550487 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/16/2010 |
| From: | Price J Dominion Nuclear Connecticut |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 10-719, FOIA/PA-2011-0115 | |
| Download: ML103550487 (11) | |
Text
Dominion Nuclear Connecticut, Inc.
5000 Dominion Boulevard, Glen Allen, Virginia 23060 Web Address: www.dom.com December 16, 2010 U. S. Nuclear Regulatory Commission Serial No.10-719 Attention: Document Control Desk NSSLIWDC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3 RELIEF REQUEST IR-3-14 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING UPDATE TO THE RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR THE THIRD 10-YEAR INSPECTION INTERVAL As part of the inservice inspection (lSI) program, Dominion Nuclear Connecticut, Inc. (DNC) submitted a letter dated March 5, 2010 requesting approval for Relief Request IR-3-14 for the update and continued implementation of a Risk-Informed Inservice Inspection (RI-ISI) program for ASME Class 1 piping at Millstone Power Station Unit 3 (MPS3). The proposed update to the lSI program, for Class 1 piping only, is based on the risk-informed methodology described in Westinghouse Owners Group WCAP-14572, Revision 1-NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report." In a letter dated November 29, 2010, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI). During an October 28, 2010 conference call between DNC and NRC, it was agreed DNC would respond to the RAI by December 21,2010.
Attachment 1 provides the DNC response to the NRC RAI addressing questions 1 through 14.
If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.
Sincerely,
Serial No.10-719 Docket No. 50-423 IR 14 Response to RAI on RI-ISI Page 2 of 2
Attachment:
- 1. Relief Request IR-3-14 Response to Request for Additional Information Regarding Update to the Risk-Informed lSI Piping for the Third 10-Year Inspection Interval Commitments made in this letter:
- 1. None cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 C. J. Sanders Project Manager - Millstone Power Station U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-883 Rockville, MD 20852-2738 NRC Senior Resident Inspector (w/o attachments)
Millstone Power Station
Serial No.10-719 Docket No. 50-423 ATTACHMENT 1 RELIEF REQUEST IR-3-14 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING UPDATE TO THE RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR THE THIRD 10-YEAR INSPECTION INTERVAL DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 1 of 8 Question 1 On page 3 of the submittal, it states: "As a result of the changes to the risk analysis, the Large Early Release Frequency (LERF) results for large and medium LOCA [loss of coolant accident] are now negligible because the Reactor Coolant System (RCS) depressurizes quickly ... " Please explain in detail the evaluation that leads to the conclusion that large and medium LOCAs depressurize so rapidly that steam generator tube rupture is prohibited. Provide the evaluation for any applicable changes to thermal hydraulic analyses and success criteria in the probabilistic risk assessment (PRA) model.
ONe Response The Millstone Power Station Unit 3 (MPS3) Level 2 probabilistic risk assessment (PRA) analysis has been updated several times since the last RI-ISI program submittal in 2000, incorporating updated thermal hydraulic analyses from cases run in the Modular Accident Analysis Program (MAAP) code. In April 2009, the current Level 2 analysis was completed using these updated th!3rmal hydraulic cases. The definition of LERF used in this analysis was the classic definition in ASME PRA Standard ASME RA-Sb-2005, that is, a large early release of fission products where large is defined as involving the rapid, unscrubbed release of airborne aerosol fission products to the environment and early is defined as occurring before the effective implementation of the off-site emergency response and protective actions. Using this definition, the only source term categories in the MPS3 Level 2 analysis that contribute to LERF are large containment isolation failures, unscrubbed intersystem LOCAs, and steam generator tube ruptures (both initiating events and induced events), which are all containment bypass terms. Quantification of the containment event trees for large and medium break LOCA events resulted in zero contribution to these four source term categories.
NUREG-1570 states that induced steam generator tube rupture is only credible for sequences in which the reactor coolant system (RCS) is at high pressure and at least one steam generator is dry. Since both large and medium LOCAs result in an RCS depressurization, they would not result in an induced tube rupture. The most likely containment failure mode for a large or medium break LOCA is containment over pressurization. Thermal hydraulic analysis for the updated Level 2 analysis shows that with containment heat removal capabilities severely degraded, it would still take over thirty hours after core damage occurs for containment to pressurize to the point of failure. This would provide sufficient time for implementation of off-site emergency response and protective actions.
This change in the Level 2 analysis had a conservative effect on the RI-ISI program.
The LERF importance of the RCS piping decreases considerably compared to the previous submittal. Because the Westinghouse RI-ISI methodology is a relative-risk based program, several high pressure safety injection segments that were not previously inspected increased in risk importance and were included in the inspection
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 2 of 8 population. This is reasonable because of the importance of high pressure safety injection in mitigating a steam generator tube rupture. No RCS piping segments were removed from the inspection population because they maintained the same importance for their contribution to core damage frequency.
Question 2 On page 3 and 4 of the submittal, a discussion on thermal cyclic fatigue was given for examination reduction of some branch lines and drain lines. For the 3 rd interval, provide the numbers of branch lines and drain lines that are being examined and the examination methods, and provide the number of lines that were examined during the 2nd interval (make a comparison between intervals).
ONC Response Specific to reduction of thermal fatigue on reactor coolant loop branch and drain piping, the following table summarizes program changes for the 3rd interval relative to the 2 nd interval. All examinations are by the ultrasonic (UT) method for the specified examination volume. No surface examinations are specified.
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Tabl e 1 Ch anges In Th erma IFf algue nspeclons na f
2 Interval 3ra Interval Location Number of R1.11 R1.20 R1.11 R1.20 Segments exams exams exams exams (Thermal (Fatigue) (Thermal (Fatigue)
Fatigue) Fatigue)
Cross Over 4 12 0 0 4 Leg Drain
[1J Hot Leg 4 12 [2J 0 8 [2J 0 Drain High Head 6 6 0 0 6 Safety Injection
[1J Residual 1 1 0 0 1 Heat Removal Suction [1]
Totals 15 31 0 8 11 Notes
- 1. These locations were determined to be not susceptible to thermal fatigue per MRP-146S.
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 3 of 8
- 2. Includes four volumetric exams of elbow body for thermal fatigue in both the 2 nd Interval and the 3rd Interval. The remaining exams are volumetric exams of welds for thermal fatigue. The reduction of one weld per segment was because the upper weld of the elbow is not required by current MRP-146 and MRP-146S requirements.
Question 3 On page 4 of the submittal, six additional high safety significant (HSS) segments resulting in five welds were added for volumetric examinations. Identify the piping systems associated with the 5 welds.
ONC Response The five additional volumetric examinations are on High Head Safety Injection (SIH) lines.
Question 4 The submittal states that the RI-ISI program is consistent with the Westinghouse Topical Report, WCAP-14572, Revision 1-NP-A with listed deviations. After the issuance of WCAP-14572, Revision 1-NP-A, the topical report was updated with later revision(s), supplement(s) and addenda. Identify the updates to WCAP-14572, Revision 1-NP-A and WCAP-14572, Revision 1-NP-A, Supplement 1, if any, that apply to the proposed alternative.
ONC Response There are no additional revisions, supplements, or addenda to WCAP-14572 that apply to the proposed alternative inspection program.
Question 5 In Attachment 1, Table 1 of the submittal, two columns are identified as "SES Matrix Region." Define the abbreviation SES and provide a discussion on what these columns are presenting.
ONC Response The abbreviation SES stands for "Structural Element Selection". The SES matrix is provided and discussed in WCAP-14572, Revision 1-NP-A, Sections 3.7.1 and 3.7.2.
Briefly, the matrix requires 100% inspection of HSS locations that are susceptible to active mechanisms (for example, cyclic thermal fatigue) and requires a statistical justification (using the Perdue Model) for the number of examinations on the remaining non-susceptible locations. Regions 1 and 2 of the SES matrix indicate "high failure importance" and "low failure importance", respectively, for HSS piping segments. The
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 4 of 8 SES column in the summary table is consistent with the standard format included in Nuclear Energy Institute guideline NEI 04-05, "Living Program to Maintain Risk Informed Inservice Inspection Programs For Nuclear Power Plant Piping Systems,"
April 2004.
Question 6 In Attachment 1, Table 1 of the submittal, the totals are listed and identified as NDE (nondestructive examinations) for the 2nd and 3rd intervals. Provide NDE values shown as totals by volumetric only, volumetric and surface, and surface only.
ONC Response The breakdown of the total non-destructive examinations (NDE) listed in Table 1 of the submittal are shown below.
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Table 2 - Changes*In RCS B J InspecfIons Examination Interval 2 Interval 3 Volumetric only 78 77 Surface only 1 0 Volumetric 0 0
+Surface Question 7 In Attachment 1, Table 1 of the submittal, the RCS lists 59 examination locations for the 2nd interval and 46 examination locations + one visual location for the 3rd interval.
Provide an explanation for the decrease in examination locations between the two intervals.
ONC Response The reduction in the number of RCS B-J examinations is a reduction of 13 NDE examinations and an addition of one visual examination. The reduction in NDE locations includes:
- the net reduction of 12 former thermal fatigue locations as shown in, Table 1, "Changes in Thermal Fatigue Inspections" (provided in the Question 2 response),
- a reduction by 2 in the number of safe-end-to-pipe welds at locations on the pressurizer that are now full structural overlays (such that the weld overlay examination now governs),
- the replacement of two B-F examinations at the steam generator Cross Over Leg nozzle with B-J examinations in the same legs (net addition of two B-J welds), and
- deletion of one B-J surface examination at a sockolet-to-pipe branch weld.
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 5 of 8 Question 8 The introduction of WCAP-14572, Revision 1-NP-A references Code Case N-577 (N-577). The NRC listed N-577 in Regulatory Guide 1.193, "ASME Code Cases Not Approved for Use." The NRC staff has accepted licensees' referencing N-577, Table 1 in submittals requesting relief from selected requirements of the ASME Code. If sections, other than Table 1 in N-577 are used in the proposed alternative, provide a description of the sections, their application, and technical justification.
ONe Response Although Code Case N-577 (N-577) is referenced in WCAP 14572, Revision 1-NP-A as a contemporary document, the WCAP-14572 methodology as approved by the NRC does not require its use. The code case was not referenced or explicitly used in ONC's analysis or preparation of the RI-ISI submittal. WCAP-14572, Revision 1-NP-A was used in a manner consistent with the safety evaluation report included with the WCAP.
Question 9 A major step in the WCAP-14572, Revision 1-NP-A, process is assignment of segments into safety significance categories based on integrated decision making process, and the selection of segments for inspection locations. The requested table summarizes the results of the safety significance categorization process as determined by the quantitative criteria and by the expert panel's deliberations based on other considerations. The summarizing information requested in the table below will provide an overview of the distribution of the safety significance of the segments based on the quantitative results, and the final distribution based on the integrated decision making.
Each segment has four risk reduced worth's (RRWs) calculated, a core damage frequency (COF) with and without operator action, and a LERF with and without operator acf Ion. PI ease provi*d e th e f 0 l i t a bl e:
OWInQ System Number of Number of Number of Number of Number of Total (Note 1) Segments Segments Segments Segments segments number of with any with any with all with any with all segments RRW> RRW RRW< RRW RRW< selected for 1.005 Between 1.001 between 1.001 inspection 1.005 and 1.005 and selected (high safety 1.001 1.001 for significant placed in inspection segments)
HHS NOTE: (1) RCS-Reactor Coolant System, SIH-High Pressure Safety Injection System, SIL-Low Pressure Safety Injection System, CHS-Chemical Volume & Control System, RHS-Residual Heat Removal System
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 6 of 8 ONCReSDonse Number of Total Number of Segments Number of number of Number Segments Number with any segments segments of with Any of RRW with all selected for Segments RRW Segments between RRW< inspection with any Between with all 1.005 and 1.001 (high safety System RRW> 1.005 and RRW< 1.001 placed selected for significance (Note 1) 1.005 1.001 1.001 in HSS inspection segments)
RCS 43 20 32 12 4 59 SIH 5 0 2 0 0 5 SIL 0 4 2 0 0 0 CHS 4 0 0 0 0 4 RHS 0 0 2 0 2 2 Total 52 24 38 12 6 70 NOTE: (1) RCS - Reactor Coolant System, SIH - High Pressure Safety Injection System, SIL - Low Pressure Safety Injection System, CHS - Chemical Volume & Control System, RHS - Residual Heat Removal System Question 10 Please verify that a sensitivity study was conducted to address uncertainty as described on page 125 (Section 3.6.1) of WCAP 14572, Revision 1-NP-A. Identify how many segments' RRW increased from below 1.001 to greater than or equal to 1.005 based on this study.
ONC Response An uncertainty analysis was performed as described on page 125 (Section 3.6.1) of WCAP 14572, Rev. 1-NP-A. The results of the uncertainty analysis were presented to the expert panel for consideration. As a result of the uncertainty analysis, no segments' RRW increased from below 1.001 to greater than or equal to 1.005.
Question 11 Please state that the change in risk calculations were performed according to all the guidelines provided on page 213 (Section 4.4.2) of WCAP 14572, Revision 1-NP-A or provide a description and justification of any deviation.
ONC Response The change in risk calculations were performed according to the applicable guidelines provided in Section 4.4.2 of the WCAP with one deviation. The scope of the proposed program for MPS3 is restricted to Class 1 piping. Class 1 piping includes the RCS
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 7 of 8 piping and small portions of other systems directly connected to the RCS piping. The justification for this deviation is that this piping is within the containment and subject to leak detection from equipment such as radiation monitors and sump level indicators.
Bullet 3 on page 213 of the WCAP discusses only RCS piping, but does not include portions of other systems connected directly to the RCS piping. For the segments included in the RCS pressure boundary that were being inspected by NDE, the failure probability with lSI was used, along with credit for leak detection. For the segments included in the RCS pressure boundary that were not being inspected, the failure probability without lSI was used, along with leak detection.
Question 12 Please provide a brief summary of qualifications possessed by the representatives on the expert panel and reasons for excluding any particular expertise mentioned for the expert panel in WCAP-14572, Revision 1-NP-A.
ONe Response The members of the expert panel completed training in the Westinghouse RI-ISI program prior to participating in the expert panel. Most members had also participated in previous expert panels for the RI-ISI program. No expertise mentioned for the expert panel in WCAP 14572, Revision 1-NP-A was excluded. As documented in the Expert Panel Meeting Minutes, the Expert Panel included representatives with expertise in system and component reliability, Maintenance Rule, probabilistic risk assessment (with experience in safety analysis), nondestructive examination, inservice inspection, plant operations, ASME code, materials and welding (with experience in pipe maintenance), piping, and engineering mechanics. This satisfies the requirements of WCAP 14572, Revision 1-NP-A.
Question 13 of DNC's request states that the MPS3 probabilistic risk assessment PRA model and documentation have been maintained as a living program and the PRA is routinely updated approximately every 3 years to reflect the current plant configuration and to reflect the accumulation of additional plant operating history and configuration failure data. Please state how DNC plans to maintain a living RI-ISI program which can be affected by changes in plant design or operations during the third 10-year interval.
ONe Response Westinghouse published LTR-RRA-05-58 Revision 1 "Risk-Informed Inservice Inspection (RI-ISI) Living Program Guidance Document" in June 2005. This document provides guidance for maintaining a living RI-ISI program which incorporates changes in plant design and operations as well as PRA changes during the program's ten year interval. The document describes periodic reviews and updates as necessary each interval to coincide with the inspection program requirements contained in ASME Section XI, Inspection Program B. The review schedule of LTR-RRA-05-58, Revision1
Serial No.10-719 Docket No. 50-423 IR-3-14 Response to RAI on RI-ISI Attachment 1, Page 8 of 8 may not coincide with the schedule for updating the PRA, but the LTR-RRA-05-58, Revision 1 reviews require that PRA changes that have taken place since the last RI-lSI program update be evaluated. During these reviews, changes which may affect the RI-ISI program are examined to determine their potential impact. If the changes identified are significant enough to affect the RI-ISI program, the program is updated accordingly to ensure it reflects the as-built, as-operated plant. DNe will continue to use the guidance in LTR-RRA-05-58 Revision 1 to review and update the RI-ISI program every period during the 3rd ten year interval.