ML040770844
| ML040770844 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/11/2004 |
| From: | Nerses V NRC/NRR/DLPM/LPD1 |
| To: | Darrell Roberts NRC/NRR/DLPM/LPD1 |
| Nerses V, NRR//DLPM, 415-1484 | |
| References | |
| TAC MC1284 | |
| Download: ML040770844 (6) | |
Text
March 11, 2004 MEMORANDUM TO: Darrell J. Roberts, Acting Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation FROM: Victor Nerses, Sr. Project Manager, Section 2 /RA/
Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 2, FACSIMILE TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (RAI) TO BE DISCUSSED IN AN UPCOMING CONFERENCE CALL (TAC NO. MC1284)
The attached draft RAI was transmitted by facsimile on March 11, 2003, to Mr. David Dodson, Dominion Nuclear Connecticut, Inc. (licensee). This draft RAI was transmitted to facilitate the technical review being conducted by NRR and to support a conference call with the licensee to discuss the RAI. The RAI was related to the licensees submittal dated November 10, 2003, concerning the implementation of a risk-informed inservice inspection program plan. Review of the RAI would allow the licensee to determine and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not convey or represent an NRC staff position regarding the licensees request.
Docket No. 50-336
Attachment:
Draft RAI
March 11, 2004 MEMORANDUM TO: Darrell J. Roberts, Acting Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation FROM: Victor Nerses, Sr. Project Manager, Section 2 /RA/
Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 2, FACSIMILE TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (RAI) TO BE DISCUSSED IN AN UPCOMING CONFERENCE CALL (TAC NO. MC1284)
The attached draft RAI was transmitted by facsimile on March 11, 2003, to Mr. David Dodson, Dominion Nuclear Connecticut, Inc. (licensee). This draft RAI was transmitted to facilitate the technical review being conducted by NRR and to support a conference call with the licensee to discuss the RAI. The RAI was related to the licensees submittal dated November 10, 2003, concerning the implementation of a risk-informed inservice inspection program plan. Review of the RAI would allow the licensee to determine and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not convey or represent an NRC staff position regarding the licensees request.
Docket No. 50-336
Attachment:
Draft RAI DISTRIBUTION PUBLIC D. Roberts V. Nerses B. Fu S, Dinsmore PDI-2 Reading Accession Number: ML040770844 OFFICE PDI-2/PM EMCB/SC SPSB/SC NAME VNerses TChan MRubin DATE 03/10/04 03/10/04 03/10/04 OFFICIAL RECORD COPY
DRAFT REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FACILITY OPERATING LICENSE NO. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 (TAC NO. MC1284)
By letter dated November 10, 2003, Dominion Nuclear Connecticut, Inc. (DNC/licensee) submitted a request to implement a risk-informed inservice inspection (RI-ISI) program plan for Millstone Power Station, Unit No. 2.
The Nuclear Regulatory Commission (NRC) staff has reviewed the information the licensee provided that supports their request. In order for the NRC staff to complete its evaluation, the following additional information is requested:
I. RR-89-40
- 1) Regulatory Guide (RG) 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking for Inservice Inspection of Piping, Revision 1, dated September 2003, replaced the original For Trial Use RG dated September 1998. Revision 1 of the RG 1.178 includes guidance on what should be included in risk informed-inservice inspection (RI-ISI) submittals, particularly in dealing with probabilistic risk assessment (PRA) issues. Specifically, on page 28 of RG 1.178, the following is stated regarding the information that should be included in a submittal:
A description of the staff and industry reviews performed on the PRA. Limitations, weakness, or improvements identified by the reviewers that could change the results of the PRA should be discussed. The resolution of the reviewer comments, or an explanation of the insensitivity of the analysis used to support the submittal to the comment, should be provided.
Your submittal briefly describes two weaknesses identified by the NRC staff during the review of the individual plant examination (IPE) and how these weaknesses have been addressed.
Your submittal also discusses a January 2000, Combustion Engineering peer review of your PRA. Please provide the Facts and Observations that the peer review team identified as important and necessary to address (Significance Level A and B in NEI 00-02 Probabilistic Risk Assessment (PRA) Peer Review Process Guidance (Rev. A3)) and describe how these issues have been resolved or why they will not affect the proposed RI-ISI program.
- 2) Did any of your segments (piping for systems that are included in the ISI program) include lengths of piping with different diameters? If some of your segments included piping of different diameters, please describe how you estimated the failure frequency of these segments and explain how this process comports with the WCAP Topical.
- 3) On page 8 of attachment 1 you state that the number of examinations in 42 of the 73 high safety significance (HSS) segments was not developed using the Perdue methodology. You further stated that, [f]or these 42 segments, the guidance in Section 3.7.3 of WCAP-14572, A-version was followed. Section 3.7.3 provides guidance on selecting inspection locations once the number of locations has been determined. Please explain how you determined the number of inspection locations for the 42 segments for which the Perdue method was not applied.
- 4) The Summary Statement at the end of Table 5-1 states, Current ASME Section XI selects a total of 155 non-destructive exams while the proposed RI-ISI program selects a total of 126 exams... Does the current ASME Section XI select a total of 155 non-destructive exams in the full population of Class 1 non-exempt welds, or from the population of non-exempt welds in the 73 HSS segments? If the current ASME Section XI selects a total of 155 non-destructive exams from the population of non-exempt welds in the 73 HSS segments, how many non-exempt welds are in the full Class 1 population and how many ASME Section XI exams are selected from this population?
- 5) In Table 3.4-1 "Failure Probability Estimates (without ISI)," please explain why stress corrosion cracking (SCC), thermal fatigue and vibration fatigue are not addressed as potential failure mechanisms for the Chemical and Volume Control System and High/Low Pressure Safety Injection systems. How will the failure probability be affected when they are considered as potential degradation mechanisms?
- 6) There has been extensive industry experience concerning cracking of alloy 600 weld materials (Inconel 82/182) in the form of primary water stress corrosion cracking (PWSCC) degradation mechanism. This degradation mechanism has not been addressed in the Topical Report WCAP-14572, Rev 1-NP-A. In Table 5-1, Structural Element Selection, 95 welds are selected for volumetric examination in B-F examination category in the reactor coolant (RC) system. Are these welds made of Inconel 82/182? Please explain how PWSCC is addressed in your program.
- 7) Under what conditions would the RI-ISI program be resubmitted to the NRC prior to the end of any 10-year interval?
- 8) Section 3.8 of the licensees submittal addresses additional examinations. It states, 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 elements initially required to be inspected on the segment or segments. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined. 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 degradation mechanism.
ASME Code directs licensees to perform these sample expansions in the current outage.
Confirm that the sample expansions of elements identified as being susceptible to the same service related root cause conditions or degradation mechanism will be completed during the outage that identified the flaws or relevant conditions.
II. RR-89-41
- 1. DNC stated in the submittal that VT-2 will be performed in lieu of volumetric examination for socket welds of any size and branch pipe connection welds of NPS 2 or smaller. What is the largest size of the socket welds?
- 2. DNC also stated that use of a volumetric examination would not provide any meaningful results... and that the use of the alternative (VT-2) provides an acceptable level of quality and safety. Please explain how a VT-2 examination can provide meaningful results. Please also explain if other non-destructive examination methods have been considered as an alternative which may provide more meaningful results than VT-2.