NL-17-1848, Systematic Risk-Informed Assessment of Debris Technical Report, SNC Response to NRC Request for Additional Information (Rals 1-3)
| ML17314A014 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/09/2017 |
| From: | Hutto J Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-17-1848 | |
| Download: ML17314A014 (6) | |
Text
A Southern Nuclear NOV 0 9 2017 Docket Nos.: 50-424 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001
). ). Hutto Regulatory Affairs Director Vogtle Electric Generating Plant-Units 1 &2 40 Inverness Center Parkway Po;t Office Box 1295 Birmingham. AL J5242 205 992 5872 tel 205 992 7601 fax jjhuno@;outhcrncu.com NL-17-1848 Systematic Risk-Informed Assessment of Debris Technical Report SNC Response to NRC Request for Additional Information (RAis #1-3)
Ladies and Gentlemen:
By letter dated April 21, 2017 (Agencywide Documents Access and Management System Accession No. ML 17116Ab96) as supplemented by letter dated July 11, 2017 (ADAMS Accession No. ML17192A245), Southern Nuclear Operating Company, Inc. (SNC) submitted a plant-specific technical report for Vogtle Electric Generating Plant, Units 1 and 2 and requested U.S. Nuclear Regulatory Commission (NRC) approval of the methods and inputs described in the technical report. The plant-specific technical report describes a risk-informed methodology to evaluate debris effects with the exception of in-vessel fiber limits. By letter dated October 12, 2017, the NRC staff notified SNC that additional information is needed for the staff to complete their review. The Enclosure provides the SNC response to the NRC requests for additional information.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.
Respectfully submitted, J. J. Hutto Regulatory Affairs Director
U.S. Nuclear Regulatory Commission NL-17-1848 Page2 JJH/PDB/CBG
Enclosure:
SNC Response to NRC Request for Additional Information (RAis)
Cc: Regional Administrator NRR Project Manager-Vogtle 1 & 2 Senior Resident Inspector-Vogtle 1 & 2 RType: CVC7000
Vogtle Electric Generating Plant Unit 1 and 2 Systematic Risk-Informed Assessment of Debris Technical Report SNC Response to NRC Request for Additional Information (RAis #1-3)
Enclosure SNC Response to NRC Request for Additional Information (RAis)
Enclosure to NL-17-1848 SNC Response to NRC Request for Additional Information (RAis)
NRC RAI1 (1)
In Table 3-9 on pages E3-40 to E3-59 of Enclosure 3 to the letter dated April 21, 2017, the licensee provides a list of information for all Class 1 welds that are considered in the GSI-191 analysis. On Pages E3-42 and E3-43, the licensee states that 4 cold leg welds are assumed to have primary water stress corrosion cracking (PWSCC) degradation mechanism in the sump analysis. Please provide an explanation as to why:
- b. The 4 cold leg welds resulted in no sump failures even though the cold leg welds are assumed to have PWSCC which will lead to weld failures.
- a. All the ASME Code Section XI Category 8-F welds nominal pipe size (NPS) 4 inch or larger and welded with Alloy 82/182 filler material were regarded as susceptible to PWSCC unless a mitigation process has been applied. The 8-F welds on the hot legs near the reactor pressure vessel (RPV) have been mitigated for PWSCC by application of mechanical devices that reverse the stress fields. In addition, stainless steel safe ends have been used to eliminate susceptibility to PWSCC for the 8-F welds on the hot legs and cold legs at the steam generators. The only 8-F welds NPS 4 inch or greater that have not been mitigated for PWSCC are four cold leg welds near the RPV.
- b. The degradation mechanisms for a weld directly affect the likelihood of a break on that specific weld (i.e., the LOCA frequency at that location). However, the GSI-191 failures shown in Table 3-9 are based on the quantity of debris generated for breaks at each weld location (i.e., assuming the breaks occur without considering the frequency). The four cold leg welds that have not been mitigated for PWSCC are in the reactor cavity, and generate significantly less debris than other break locations. The debris quantities generated at these four locations are sufficiently low that none of these breaks exceed any of the GSI-191 acceptance criteria.
NRC RAI2
{2)
In Section 2.3.2 on page E4-11 of Enclosure 4 to the letter dated April 21, 2017, the licensee states, in part, that:
The inservice inspection (lSI) program provides rules for the examination and repair of piping and other RCS [reactor coolant system] components, and plays an important role in the prevention of pipe breaks. The integrity of the Class 1 welds, piping, and components are maintained at a high level of reliability through the ASME Section XI inspection program [the 2010 edition,] (Reference 6).
VEGP lSI procedures also ensure that inspections are performed in accordance with the schedule requirements of the code.
- a. Please identify the welds that are mitigated to minimize PWSCC and the associated mitigation method(s) for those welds.
E-1
Enclosure to NL-17-1848 SNC Response to NRC Request for Additional Information (RAis)
- b.
Please identify the Alloy 82/182 welds that have not been mitigated.
- c. Please discuss any flaws or indications that have been detected, but not repaired, in any of the welds that are considered in the GSI-191 evaluation.
- a. As discussed in the response to RAI 1.a, the 8-F welds on the hot legs near the RPV have been mitigated for PWSCC by application of mechanical devices that reverse the stress fields. In addition, stainless steel safe ends have been used to eliminate susceptibility to PWSCC for the 8-F welds on the hot legs and cold legs at the steam generators.
All 8-F welds attached to the pressurizer (including the surge line and pressurizer spray and relief valve piping) have been mitigated by application of a full structural weld overlay.
- b. The only Category 8-F NPS 4 inch or larger Alloy 82/182 welds that have not been mitigated are the four cold leg welds discussed in RAI 1 :
Weld 11201-V6-001-W35-R8 Weld 11201-V6-001-W38-R8 Weld 11201-V6-001-W34-R8 Weld 11201-V6-001-W39-R8 Other locations with unmitigated 82/182 welds are listed below. However, this set of welds is not included in the debris effects model because these welds are not RCS piping welds:
SG channel head drain (managed under Code Case N-722-1)
RV head CRDMs (managed under Code Case N-729-4)
RV 8Mis (managed under Code Case N-722-1)
RV head vent, (managed under Code Case N-729-4)
- c. No flaws have been detected in any of the unmitigated Alloy 82/182 welds NRC RAI3 (3)
In Section 2.3.2 on page E4-11 of Enclosure 4 to the letter dated April 21, 2017, the licensee states, in part, that:
The leak detection program at VEGP is capable of early identification of RCS leakage in accordance with RG [Regulatory Guide] 1.45, Revision 1, ["Guidance on Monitoring and Responding to Reactor Coolant System Leakage," May 2008,]
(Reference 8) to provide time for appropriate operator action to identify and address RCS leakage...
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Enclosure to NL-17-1848 SNC Response to NRC Request for Additional Information (RAis)
Please discuss any changes or enhancement, if any, to the RCS leakage detection systems after Revision 16 of Chapter 5.0, "Reactor Coolant System and Connected Systems," of the VEGP Updated Final Safety Analysis Report was provided to the NRC.
SNC Response to RAI 3 The VEGP RCS leakage detection system is described in FSAR Section 5.2.5. There have been no changes or enhancements to the system since this section was last updated in April 2015.
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