ML102570807
| ML102570807 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/13/2010 |
| From: | Price J Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 10-050A | |
| Download: ML102570807 (23) | |
Text
10 CFR 50.55a VIRGINIA E L E C T R I C AND POWER C O MPAN Y RICHMOND, VIRGINIA 23261 September 13, 2010 U.S. Nuclear Regulatory Commission Serial No. 10-050A Attention: Document Control Desk NL&OS/ETS RO Washington, D.C. 20555 Docket Nos. 50-338/339 License Nos. NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
NORTH ANNA POWER STATION UNITS 1 AND 2 ASME SECTION XI INSERVICE INSPECTION PROGRAM RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST N1-14-RI-001 AND N2-14-RI-001 REQUEST FOR ALTERNATIVE -IMPLEMENTATION OF A RISK-INFORMED INSERVICE INSPECTION PROGRAM BASED ON ASME CODE CASE N-716 In a February 23, 2010 letter (Serial No.10-050), Dominion requested authorization to implement a risk-informed inservice inspection (RI-ISI) program based on the American Society of Mechanical Engineers (ASME) Code Case N-716, as documented in the Requests for Alternative N1-14-RI-001 and N2-14-RI-001 for Units 1 and 2, respectively.
N1-14-RI-001 and N2-14-RI-001 were submitted in a template format. In an August 12, 2010 e-mail from Dr. V. Sreenivas, the NRC requested additional information to complete the review of the RI-ISI program. The attachment to this letter provides the requested information .
Dominion plans to implement this alternative for the entire 4th lSI Interval for North Anna Units 1 and 2. North Anna Unit 1's 4th 10-Year Interval began May 1, 2009 and will end April 30,2019. North Anna Unit 2's 4th 10-Year Interval begins December 14, 2010 and will end December 13, 2020. Therefore, Dominion continues to request review and approval of N1-14-RI-001 and N2-14-RI-001 by February, 2011 in order to plan and complete the first period examinations.
If you have any questions or require additional information , please contact Mr. Thomas Shaub at (804) 273-2763.
Respectfully, J. a Price Vipe resident - Nuclear Engineering Attachment
- 1. Response to Request for Additional Information - Relief Requests N1-14-RI-001 and N2-14-RI-001 with 3 Enclosures
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 2 of 2 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue, NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. J. E. Reasor, Jr.
Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.
Suite 300 Glen Allen, Virginia 23060 NRC Senior Resident Inspector North Anna Power Station Ms. K. R. Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Attachment RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION INSERVICE INSPECTION PLAN RISK INFORMED FOURTH INTERVAL RI RELIEF REQUESTS N1-14-RI-001 AND N2-14-RI-001 North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 1 of 13
Background
By letter dated February 23, 2010, Virginia Electric and Power Company (Dominion),
submitted for staff review and approval lSI Program Relief Requests N1-14-RI-001 and N2-14-RI-001, which request approval to use alternative risk-informed inservice inspection (RI-ISI) selection and examination criteria for Category B-F, B-J, C-F-1 and C-F-2 pressure retaining piping welds for the North Anna Power Station (NAPS) Units 1
& 2. To complete their review, the NRC staff requested the following additional information.
Probabilistic Risk Assessment Licensing Branch The NRC has not endorsed EPRI (Electric Power Research Institute) Topical Report 1018427. The questions listed below address the quality of North Anna probabilistic risk assessment (PRA) model.
Question 1 .
A self assessment performed on the North Anna PRA model in August 2007 identified PRA modeling and documentation supporting requirements (SRs) where the PRA model did not meet Capability Category (CC) " of the ASME (American Society of Mechanical Engineers) PRA standard. In December 2009, a model update was performed to meet Category" of the ASME PRA standard and Regulatory Guide 1.200 Rev 1. Please identify and disposition any remaining differences with CC "
requirements (i.e., open items) that may affect this application.
Dominion Response A self assessment was performed on the North Anna PRA model in June 2010 identifying remaining open items that do not meet Regulatory Guide 1.200 Rev 1 Capability Category (CC) II requirements. These open items are identified and evaluated for their potential to impact the risk assessment performed for the RI-ISI program at North Anna. Specifically, the unmet supporting requirements (SRs) are considered for their ability to impact the quantification of a large break LOCA, which was the bounding case that was used for the change in risk analysis, or the internal flooding analysis, which was used for scope determination. Based on the evaluation in the table below, none of the open items identified in the current North Anna PRA model affect the inputs or results of this application.
Ser ial No. 10-050A Docket Nos. 50-338/339 Respo nse to Request for Additional Information Fou rth Interval Risk Informed Relief Requests N1-1 4-RI-001 & N2-14-R I-001 Page 2 of 13 Gap Description Self Assessment Impact on the RI-ISI Applicat ion AS-A 4 For key safety func tions SR remains as NOT MET until 1) The importance of the SBO Diesel (e.g., power restorat ion) 1) an huma n event probability is low with respect to floodi ng events identify operator actio ns (HEP) is added to the station and a large break LOGA event in the to achieve the defined Blackout (SBO) nodes for North Anna PRA mode l. The Risk success criteria. resto ring the EGGS func tions ; Ach ievement Worth (RAW ) of the and 2) text in section 2.3.3.1 is SBO for flooding events is 1.00, and revised to clarify the need for the SBO RAW for a large break operator action to restart EGGS LOGA event is 1.01. Based on this funct ions. low risk worth, adding an HEP to the SBO nodes for restoring the EGGS fu nctions would not impact the results of the risk assessment performed to support this application.
- 2) Not Significant. This is judged to be a documentation consideration only and does not affec t the technical adequacy of the PRA mode l.
AS-B5a Define and model plant The NAPS models credit use of For all of the crosst ie systems , either configu rations and the oppos ite unit systems, e.g., the unavailability during refueling alignments that reflect charging system and diesel- outages is accounted for in the PRA dependencies. generators, for accident unavailabilities or the system/trains mitigation. However, no do not have significant unavailability documentation was identified during outages. The only except ion that would show how oppos ite are the electrical buses where the unit outages were considered. unavailab ility during at power For example, during a refueling operat ion is not included in the PRA outage , a Train-A outage may model. The estimated unavailability make charg ing or component is 1-2 days , which is less than 1E-2 cooling (GC) cross-t ie change in estimated unava ilability.
unavailable for a significant Th is small change in unavailability period of time. Such would not impact the flooding unava ilability values could evaluat ion or the quantification of a reach 5% overal l. If large break LOGA.
unavailability during opposite-unit outages is included in the overall system unava ilability, then that could be stated in the accident seque nce (AS) documentation.
DA-D2 W hen using expert Documentation needs to be Not Signif icant. This is judged to be judgment document the enhanced for the several cases a documentation consideration only rationa le behind the where expert opinion is used. and does not affect the techn ical choice of parameter The expe rt opinion is adequacy of the PRA model.
values. reasonable and should not chance.
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Addit ional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 3 of 13 Gap Description Self Assessment Impact on the RI-ISI Applicat ion QU-B1 Identify method-specific Although key assumptions are Not Significant. This is judged to be limitat ions and features documented, these do not a documentation consideration only that could impact the include limitations of the and does not affect the technical results and applications. quantification method or adequacy of the PRA model.
features that impact results (aside from references to code limitations, guidance documents and procedures).
QU-F5 Identify method-specific Although key assumptions are Not Significant. This is judged to be limitations and features documented, these do not a documentation consideration only that could impact the include limitations of the and does not affect the technical results and applications. quantif ication method or adequacy of the PRA model.
features that impact results (aside from references to code limitations, guidance documents and procedures).
SC-A6 Include a discussion of Some of the success criteria Not Significant. This is judged to be operator actions discussion includes general a documentation consideration only assumed as part of the operator actions, but the and does not affect the technical success criteria discussion does not include adequacy of the PRA model.
development, and how procedures and not all event those actions are tree sections contain the consistent with plant discussion.
procedures and practices.
SY-A2 Use results of plant The Dominion PRA staff has Not Sign ificant. This is judged to be walkdowns and plant performed many system a documentation cons ideration only personnel interviews walkdowns during the and does not affect the technical (system engineers and development and maintenance adequacy of the PRA model.
operators) as a source of of the models . In addition ,
information for modeling Dominion PRA staff works the as-bui lt, as-operated closely with North Anna system plant. engineers and operators on nearly a daily basis while supporting the various risk informed programs. However, no formal documentation exists at this time to allow closure of these supporting requirements (SRs). It is NOT anticipated that not meeting this requirement will have a siqnificant impact on the model.
SY-B15 Identify SSCs that may Currently, the NAPS PRA Including a specific failure probability be required to operate in model does not distingu ish for a pressurizer PORV failing to condit ions beyond the ir between PZR PORVs failing to reseat after passing water would not environmental reclose on water or steam impact the internal flooding qualifications. relief. See EPRI TR-1011047 evaluation or large break LOCA "Probability of Safety Valve quantification, so this open item does Failure-to-Reseat Following not impact the RI-ISI application .
Steam and Liquid Relief."
Serial NO.1 0-050A Docket Nos. 50-338/339 Response to Request for Add itional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 4 of 13 Gap Description Self Assessment Impact on the RI-ISI Application SY-B8 Use results of plant The Dominion PRA staff has Not Significant. Th is is judged to be walkdowns and plant performed many system a documentation consideration only personnel interviews walkdowns during the and does not affect the techn ical (system engineers and development and maintenance adequacy of the PRA model.
operators) as a source of of the models. In addition, information for modeling Dominion PRA staff works the as-bu ilt, as-operated closely with North Anna system plant. engineers and operators on nearly a daily basis while supporting the various risk informed programs. However, no formal documentation exists at this time to allow closure of these SRs. It is NOT anticipated that not meeting this requirement will have a significant impact on the model.
HR-G4 Base the time available Time windows for successful As part of the 2009 model update, to complete actions on completion of act ions in some new MAAP runs were performed for appropriate realistic instances may need to be some of the key operator act ions.
generic thermal - updated (for example, those This includes runs to support the hydraulic ana lyses, or that are based on estimates est imation of HEP -1 FRH:1-11 and simulation from similar made for the IPE). HEP-1 FRH 1-15-0NE, which are the plants only HEPs that could significantly affect the internal flood ing evaluation. The HEP for transferring to hotleg recirculation , HEP -1ES1 :4, is the only HEP important for large break LOCA quantification.
Updating the timing for addit ional HEPs to meet CC II of this SR would not im pact the risk assessment performed for the RI-ISI application .
HR-G5 Base the required time No documentation cu rrently Not Significant. This is judged to be to complete actions fo r exists and this SR will remain a documentation consideration only significant HFEs on NOT MET. As a footnote the and does not affect the techn ical action time timings are not expected to adequacy of the PRA model.
measurements in either change significantly as they are walkthroughs or talk- based on comparisons with throughs of the similar actions at Surry.
procedures or simulator observations.
SY -A4 Use resu lts of plant The Dominion PRA staff has Not Significant. This is judged to be walkdowns and plant performed many system a documentation consideration only personnel interviews walkdowns during the and does not affect the techn ical (system engineers and development and maintenance adequacy of the PRA model.
operators) as a source of of the models. In addition, information for modeling Dominion PRA staff works the as-built, as-operated closely with North Anna system plant. engineers and operators on nearly a daily bas is while
Serial No.1 0-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1- 14-RI-001 & N2-14-RI-001 Page 5 of 13 Gap Description Self Assessment Impact on the RI-ISI Application supporting the various risk informed prog rams. However, no formal documentation exists at this time to allow closure of these SRs . It is NOT anticipated that not meeting this requi rement will have a significant impact on the model.
AS-A7 Delineate accident SR is NOT MET until : 1) 1) Inclusion of consequent ial RCP sequence (e.g., Loss of inclusion of consequential loss seal cooling for transients would not RCP seal cooling) for of RCP seal cooling for affect the appl icat ion because only each initiating event transients, and 2) the large break LOCA and flooding (e.g., transients). documentation enhancement of events were quantified.
the U1-RCPSL nodes. Consequential loss of RCP cooli ng is considered for flooding events in the North Anna PRA model , and does not apply for a large break LOCA scenario. 2) Not Significant. This is judged to be a documentation consideration only and does not affect the technical adequacy of the PRA model.
QU-E1 Identify key sources of Each PRA element notebook Not Sign ificant. This is judged to be model uncertainty. (IE , AS, SC, SY, DA, HR, LE) a documentation consideration only has identified potential sources and does not affect the technical of model uncertainty. A adequacy of the PRA model.
characterization of those sou rces of uncertainty and evaluation of the generic sources of uncertainty has not yet been completed however.
QU -F4 DOCUMENT key Although the different element Not Significant. This is judged to be assumptions and key notebooks (IE , AS , SC , SY , a documentation consideration only sources of uncertainty, etc.) do include specific and does not affect the techn ical such as: possible assumptions related to the adequacy of the PRA model.
optimistic or development of that element, conservative success there is no discussion in the criteria , suitability of the QU.1 (input) and QU.2 (results) reliability data, possible notebooks of the sources of modeling uncertainties uncertainty in the NAPS model, (model ing limitations due nor of the assumptions to the method selected), associated with those degree of completeness uncertainties.
in the selection of initiating events, possible spatial dependencies, etc.
Serial No. 10-050A Docket Nos. 50-338/339 Response to Reques t for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 6 of 13 Gap Desc ription Self Assessment Impact on the RI-ISI Application LE-D4 PERFORM a realistic Secondary side isolation is Not Signif icant. This is with regards secondary side isolation explicitly and realistically to SGTR initiating event and would capab ility analysis for modeled in the Level 1 System not impact the flood ing evaluation or the significant accident Analysis notebooks for pre-core the quantification of a large break progression sequences damage cons ideration. LOCA.
caused by SG tube However, secondary side release . USE a isolation during a SGTR should conservative or a also consider the additional comb ination of number of demands on the conservative and relief valves in the progression realistic evaluation of to core damage. It is possible secondary side isolation that some sequences capab ility for non- considered "isolated" in the signif icant accident Level 1 analysis could be progression sequences unisolated in the Level 2 result ing in a large early analysis. Also , vers ion 4 of the release . JUSTIFY MAAP code provides better applicability to the plant SGTR analysis than had been being evaluated. used for the IPE with version 3 Analyses may co nsider of the code.
realistic comparison with similar isolation capability in similar containment designs.
QU-E2 IDENTIFY key The QU.1 (input) notebook Not Significant. As part of the 2009 assumptions made in the indicates that key modeling model update, within each PRA development of the PRA assumptions are documented in element notebook (IE, AS, SC, SY, model. Part II of the PRA model DA, HR, LE) , potential sources of notebook, but this part has not model uncertainty have been yet been developed (although identified. A characterization of those some key assumptions may be sources of uncertainty and available in the IPE submittal), evaluation of the generic sources of The different element uncertainty has not yet been notebooks (IE, AS, SC, SY, completed however. This is judged etc.) do include specif ic to be a documentation consideration assumptions related to the only and does not affect the development of that element, technical adequacy of the PRA but there is typically no model.
discussion of the sources of uncertainty those assumptions relate to and the impacts of those assumptions.
QU-E3 ESTIMATE the The QU.1 (input) and QU.2 Not Significant. The parametric uncertainty interval of (resul ts) notebooks do not uncertainty analysis has been the overall CDF results . include a parametric uncertainty drafted and documented in notebook ESTIMATE the analysis. Although QU.1 does QU.3, wh ich is currently undergoing uncertainty intervals note that the basis event data acceptance review. The parametric associated with (BED) file contains uncertainty uncertainty analysis has been parameter uncertainties distribution data and the basic performed with correlated basic (DA-D3, HR-D6, HR-G9, event uncertainty data in the events in order to reflect "state-of-IEC13) , taking into parameter file is documented in knowledge" dependencies. This is
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 7 of 13 Gap Description Self Assessment Impact on the RI-ISI Application account the "state-of- the data notebooks (section judged to be a documentation knowledge " correlation . 2.5), and that uncertainty cons ideration only and does not analyses can be performed on affect the technical adequacy of the the equation files (section 4.0) , PRA model.
there is no such analysis mentioned in QU.2 .
There are a few basic events in the parameter file (N05A_16C .prm) that do not contain uncertainty distribution data.
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Addit ional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 8 of 13 Question 2 The supporting requirement (SR), IF-C6 and IF-C8, permits screening out of flood areas based on, in part, the success of human actions to isolate and terminate the flood. The endorsed RI-ISI methods require determination of the flood scenario with and without human intervention which corresponds to the capability category /1/, i.e., scenarios are not screened out based on human actions. Therefore a category /1/ analysis would be acceptable. To provide confidence that scenarios that might exceed the quantitative CDF and LERF guideline are identified, please describe how credit is given to human actions if the current application analysis does not meet Capability Category /1/ for these supporting requirements.
Dominion Response Floods were not screened out based on the ability of human actions to isolate or mitigate a flood in the North Anna PRA model. The model meets Capability Category III for IF-C6 and IF-C8, which is appropriate for this application.
NDE Branch Question 1 Table IWB-2500-1 of ASME,Section XI, 2001 Edition with 2003 Addenda requires volumetric and/or surface examination of all Category B-F or B-J Pressure Retaining Dissimilar Metal Welds greater than NPS 1. Based on recent findings of primary water stress corrosion cracking (PWSCC) in Alloy 82/182 dissimilar metal welds the staff would like more information on your inspection plans for these welds in the 4 h Interval lSI Plan for NAPS Units 1 & 2.
Describe the inspection plan of Alloy 82/182 dissimilar metal welds greater than NPS 1 in the 4th Interval lSI Plan for NAPS Units 1 and 2 (e.g., are these welds included in the number of welds selected for examination in the RI-ISI program, how many of these welds are selected for examination, what examination methodes) are being employed, what is the frequency of examination, how is disposition of limited coverage <<90%)
examinations handled, etc.).
Dominion Response PWSCC is an active degradation mechanism (OM) included in the Code Case N-716 (RIS-B) analysis. The checklist criteria for PWSCC is:
- a. piping material is Inconel (Alloy 600), and
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Informat ion Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 9 of 13
- b. exposed to primary water at temperatures greater than 570 0 F, and
- c. the material is mill-annealed and cold worked, or cold worked and welded without stress relief.
The method of examination for PWSCC susceptible welds is a volumetric examination using ASME Section XI Figure IWB-2500-8(c). The volume shall be increased by enough distance, approximately 1/2 inch, to include each side of the base metal thickness transition or counterbore transition.
Two areas at North Anna were recognized as susceptible to PWSCC in the RIS-B analysis: 1) Unit 1 Steam Generator hot leg nozzles, and 2) the pressurizer nozzle to safe-end welds, including those for the surge line, spray line and safety and relief valve lines for both units. The Unit 2 Steam Generator hot leg nozzles, which contain the Alloy 82/182 material that is susceptible to PWSCC were inlaid with Alloy 52 during preservice construction. Therefore , the Alloy 82/182 material has never been exposed to primary grade water.
For Unit 1, seven components were assigned PWSCC susceptibility and four (57%)
were selected for examination. Two Unit 1 components were assigned DMs for both PWSCC and TT (Thermal Transients). Both of these two components have been selected for examination . For Unit 2, four components were assigned PWSCC susceptibility and one (25%) was selected for examination. Two Unit 2 components were assigned both PWSCC and IT DMs and both were selected for examination. By the Code Case criteria a minimum of 25% of the DMs or combination of DMs should be selected for examination per interval. These selections will be examined once per interval. In this manner, the Class 1 Alloy 82/182 dissimilar metal welds were included in the population analyzed by the Code Case N-716 application to make component selections for examination.
All of the dissimilar metal welds on both Unit 1 and 2 pressurizers have been overlaid with PWSCC resistant material to reinforce the structural integrity. Relief Request NDE-005 for North Anna Unit 1 Interval 4 was developed to address the inspection method and frequency for the pressurizer overlays and was approved by the NRC in a letter dated September 28, 2009 (ML092530274). Dominion plans to inspect the Unit 1 overlaid welds in accordance with this relief request at this time. The same sampling selection of 25% is required by the relief request, so the number of inspections is consistent with the RIS-B approach. However, Code Case N-716 does not address welds that have been overlaid. The inspection techniques and all aspects of the relief request (including determination of additional exams and evaluating indications) will be followed for Unit 1 overlaid selected examinations.
Dominion anticipates Code Case N-770, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Informat ion Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 10 of 13 Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities" will be incorporated by the NRC into the next rule change of the Federal Register by May 2011. This new rule should be effective in time to establish the examination rules for Unit 2 pressurizer overlaid weld inspections.
If incorporation of Code Case occurs as anticipated, Dominion will then use the requirements of Code Case N-770 to govern inspection requirements for welds fabricated with Alloy 82/182 material and withdraw the Unit 1 relief request. The inspection selections determined by the RIS-B Program will remain unchanged.
However, the guidance of this Code Case will determine the examination methods. If needed, a relief request similar to Unit 1 will be submitted for the Unit 2 pressurizer overlaid Alloy 600 welds, to address inspection method and technique.
North Anna currently has an Augmented Plan that addresses MRP-139, "Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline." Examinations must be performed on Alloy 600 welds until the welds have been mitigated. Currently, the Unit 1 Steam Generator hot and cold leg nozzle to safe end welds must be examined as follows:
- One hot leg nozzle weld: Bare metal visual inspection every refueling outage, UT every period
- One cold leg nozzle weld: Bare metal visual inspection every refueling outage, UT every five years Code Case N-722 was incorporated into the last publication of the Code of Federal Regulations and was implemented by January 2009 at North Anna. The Code Case requires a visual bare metal (VE) inspection on unmitigated, Class 1, Alloy 600 welds.
Steam Generator hot leg nozzle-to-pipe-welds must receive a VE inspection every refueling outage and the Steam Generator cold leg nozzles must be VE inspected once per interval.
If any RIS-B selections are made on welds that have not been mitigated (i.e., Steam Generator hot legs) they will be volumetrically examined in accordance with Code Case N-716 using ASME Section XI Figure IWB-2500-8(c). The volume shall be increased by enough distance, approximately 1/2 inch, to include each side of the base metal thickness transition or counterbore transition. Dominion has used the phased array UT technique previously on the Steam Generator nozzle welds to achieve maximum coverage and obtain acceptable results and plans to continue to use the phased array technique on the Steam Generator welds. .
Part of the initial selection process for determining RIS-B examinations is to choose components that are known to meet full coverage requirements. If limited examinations
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 11 of 13 (coverage 90% or less) do occur, Dominion will address the limitations by a relief request in accordance with 10 CFR 50.55a(g)(5)(iii) .
To summarize, presently there are several drivers to inspect Alloy 600 welds susceptible to PWSCC: Code Case N-722, MRP-139 and the proposed Risk Informed Inservice Inspections. North Anna has programs in place to address each of these independently. The RIS-B analysis was performed without consideration of any other Programs for inspecting Alloy 600 welds; such as MRP-139 and Code Case N-722.
The analysis of the RIS-B Program did not credit exams scheduled to meet N-722 or the Augmented Program (MRP-139) in any manner to reduce the need for inspections. If welds have been overlaid they are still noted in the RIS-B Program as susceptible to PWSCC and will be selected for examination as required. The technique for examining the overlaid welds will follow criteria of either an accepted relief request or Code Case N-770 if incorporated into the Code of Federal Regulations. Examination overlap may occur and credit for one weld exam may be taken for multiple programs if the examination specifications/requirements for each program credited are met.
NRC Question 2 Section 3.3 of the February 23, 2010 submittal states that, "In contrast to a number of RI-ISI Program applications where percentage of Class 1 piping locations selected for examination has fallen substantially below 10%, Code Case N-716 mandates that 10%
be chosen." Immediately below this paragraph a brief summary is provided showing the number of welds in Class 1, 2 and non-class systems along with the number of welds selected for examination for NAPS Units 1 & 2. According to this summary the number of Class 1 welds selected for examination on Unit 2 is significantly less than 10% of the total number of Class 1 welds. Please explain this discrepancy.
Dominion Response At North Anna, the Class 1 boundaries have been unnecessarily extended beyond the second isolation valve from the reactor pressure vessel. This was done to coordinate the "Q" Quality boundary designations that were made during system design and construction.
In determining the High Safety Significant (HSS) components that are subject to selection for examination, Code Case N-716 defines HSS welds as:
(1) Class 1 portions of the reactor coolant pressure boundary (RCPB), except as provided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii).
10CFR50.55a(c) Reactor Coolant Pressure Boundary (2)(ii) states "The component is or can be isolated from the reactor coolant system by two valves in series (both
Serial NO.1 0-050A Docket Nos. 50-338/339 Response to Request for Addit ional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 12 of 13 closed, both open, or one closed and the other open). Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only."
Part (2) of the Code Case further defines HSS components:
(2) applicable portions of the shutdown cooling pressure boundary function. That is, Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:
(a) as part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (Le., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds.
Section 3.1 of the February 10, 2010 submittal reiterates this information.
Based on the definitions in Code Case N-716, 263 of Unit 2 Class 1 Safety Injection welds are Low Safety Significant (LSS), 513 of Class 1 Charging welds are LSS.
For Unit 1, 38 of Class 1 Safety Injection welds are LSS and 77 Class 1 Charging welds are LSS. These welds are not required to be included in the HSS population, but are included in the total Class 1 weld count in the table of Section 3.3 of the February 10,2010 submittal.
The statement in Section 3.3 when addressing the unique classification definitions at North Anna, would be better stated, "In contrast to a number of RI-ISI Program applications where the percentage of HSS Class 1 piping locations selected for examination has fallen substantially below 10%, Code Case N-716 mandates that 10% be chosen." At most plants the total number of Class 1 components will be HSS; however, at North Anna Units 1 and 2, they are not.
Dominion's engineering document "Risk Informed Inservice Inspection Program for NAPS 1 and 24th Intervals, Code Case N-716 Based" was written to support the RI-ISI submittal. Enclosure 1 contains pages of that Engineering document. The Weld Count tables show the LSS and HSS totals . Using the RIS-B application, 10%
of the HSS welds are required to be examined, which is 10 % of 1433 for Unit 1 and
Serial No. 10-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 13 of 13 10% of 1528 for Unit 2. The total number selected shown in the Section 3.3 table of the Submittal , 178 for Unit 1 and 183 for Unit 2 is correct.
NRC Question 3 Also the total number of welds shown in the summary in Section 3:3 for Unit 2 does not agree with the "Weld Count" column total value shown in Table 3. 1b of the February 23, 2010 submittal. Please explain this discrepancy.
Dominion Response Table 3.1 b for NAPS 2 was not correct and contained erroneous totals for Low Safety Significant welds. The following are the correct values for total LSS welds: Main Steam-- 171 versus 160, Residual Heat - 141 versus 139, Safety Injection - 736 versus 734, Quench Spray - 167 versus 165, Recirculation Spray - 88 versus 86. Total LSS welds should be 2244 versus 2234 and the total weld count should be 3772 versus 3762. Enclosure 2 to this letter contains a corrected table. Please replace Table 3.1 b in the original submittal with the updated information presented in Enclosure 2. The total weld counts and number of selections in the Section 3.3 table of the submittal is correct.
During the development of this response we discovered a typographical error. On Table 3.3a, for the RC system TT (Thermal Transients) should pair with PWSCC, not TASCS (Thermal Stratification, Cycling and Striping). Enclosure 3 is the corrected Table 3.3a for NAPS 1. Please replace this Table in the original submittal.
Serial NO.1 0-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth lnterval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Enclosure 1 North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)
ET-ISI-2010-0001 Rev. 1 NAPS 1 Code Case N-716 Selection Summary Attachment 10 pg 1 of 2 DM Welds (25%) HSS (10%) RCPB (10%) RCPBu (2/3 of RCPB ) BER(10 %)
Syste m T ota l Selected % DM To ta l HSS % T otal Selecte d 0/0 T ot al Selecte d % RCPB T ot al Selecte d % BE R CH - Charging 18 5 27.8% 16.0% 3 19 51 16.0% 59 49 96.1 % 0 0 0.0%
FW - Main Feedwater 0 0 0.0% 11.3% 0 0 0.0% 0 0 0.0% 43 10 23 .3%
MS - Main Stea m 0 0 0.0% 20.0% 0 0 0.0% 0 0 0.0% 30 6 20 .0%
RC - Reactor Coo lant 77 22 28.6% 10.9% 570 62 10.9% 452 62 100.0% 0 0 0.0 %
RH - Residual Heat 0 0 0.0% 12.9% 31 4 12.9% 9 2 50.0% 0 0 0.0%
SI - Safety Injection 9 3 33.3% 11.4% 368 42 11.4% 0 0 0.0% 0 0 0.0%
Tota l 104 30 28.8 % 178 1288 159 12.3% 520 113 7 1.1% 73 16 21.9%
Total Check 104 30 28.8% 12.4% 1288 159 12.3% 520 113 71.1% 73 16 21.9%
Tot al Selecte d 178 Total Section XI Inspe, 294 We ld Cou nt Sys te m T otal H SS LSS CH- Charging 811 319 492 FW - Main Fee dwater 115 115 0 MS - Main Steam 199 30 169 RC - Reactor Coolant 570 570 0 RH - Residual Heat 168 31 137 SI - Safety Injection 860 368 492 QS - Quench Spray 143 0 143 RS - Recirc Spray 73 0 73 Total 2939 1433 1506
ET-ISI-2010-0001 Rev. 1 NAPS 2 Code Case N-716 Selection Summary Attachme nt 10 pg 2 of 2 DM We lds (25%) HSS (10'10) RCPB (10%) R CPBu (2/3 of RCPB) BER (10%)
System Total Selected % DM Tota l HSS %. To ta l Selected % To ta l Selected % R CPB T otal Selected % BE R CH - Charging 20 6 30.00% 11.90% 420 50 11.90% 58 47 94.00% 0 FW - Feedwater 0 0 0.00% 13.27% 0 0 0.00% 0 0 0.0% 44 15 34.1%
MS - Main Steam 0 0 0.00% 18.18% 0 0 0.00% 0 0 0.0% 26 6 23.1%
RC - Reactor Coolant 78 27 34.62% 11.87% 573 68 11.87% 472 68 100.0% 0 0 0.0%
RH - Residual Heat Removal 0 0 0.00% 13.79% 29 4 13.79% 10 2 50.0% 0 0 0.0%
SI - Safety Inj ection 9 3 33.33 % 11.11% 360 40 11.11% 0 0 0.0% 0 0 0.0%
Total 107 36 33.64% 183.00 1382 162 11.72% 540 117 72.2% 70 21 30.0%
Total Check 107 36 33.64% 11.98% 1382 162 11.72% 540 117 72.2% 70 21 30.0%
Total Selected 183 Total Section XI Inspected 279 Weld Count Syste m Total HSS LSS CH - Charg ing 1340 420 920 FW - Main Feedwater 134 113 21 MS - Main Steam 204 33 17 1 RC - Reactor Coolant 573 573 0 RH - Residual Heat Removal 170 29 141 SI - Safety Injection 1096 360 736 QS - Quench Spray 167 0 167 RS - Recirc Spray 88 0 88 Tot al 3772 1528 2244
Serial No.1 0-050A Docket Nos. 50-338/339 Response to Request for Add itional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Enclosure 2 North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)10-050 Pg 16 of 26 Table 3.1b N*716 Safety Significance Determination: NAPS2 System Description Weld N-716 Safety Significance Determination Safety Count Significance RCPS SOC PWR: BER >1E_6c DF High Low FW >1E_7LERF 420 CH - Charging
" "
920
"
" "
69
" "
21
"
MS - Main Steam 33 171 " " "
RC - Reactor Coolant 573
" "
RH - Residual Heat 29 141 " " "
"
SI - Safety Injection 360
" "
736
"
QS - Quench Spray 167
"
R8- Recirculation Spray 88
"
SUMMARY
RESULTS FOR ALL SYSTEMS 1353
" "
77 69 .; " "
"
29 2244 " "
"
TOTALS 3772 1528 2244
Serial No.1 0-050A Docket Nos. 50-338/339 Response to Request for Additional Information Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Enclosure 3 North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)10-050 Pg 18 of 26 Table 3.3a N-716 Element Selections: NAPS1 System(1) Selections HSS(2) DMs(3) RCPB(4) RCPB1RV(5) RCPBOC(6j BER(7)
CH Required 32 of 319 rr 50f8 32 of 319 37 n/a n/a Made 51 n5 51 49 n/a n/a FW Required 12 of 115 n/a n/a n/a n/a 5 of 43 Made 13 n/a n/a n/a n/a 10 MS Required 3 of 30 n/a n/a n/a n/a 3 of 30 Made 6 n/a n/a n/a n/a 6 RC Required 57 of 570 TASCS, rr 4 57 of 570 41 n/a n/a TASCS 11 of 41 rr 30f11 PWSCC 20f7 PWSCC, rr 1 of 2 Made 62 TASCS, rr 7 62 62 n/a n/a TASCS 5 n4 PWSCC 4 PWSCC, rr 2 RH Required 3 of 31 n/a 3 of 31 3 n/a n/a Made 4 n/a 4 2 n/a n/a SI Required 37 of 368 IGSCC 2 of 6 37 of 368 0 n/a n/a rr, IGSCC 1 of 3 Made 42 IGSCC 2 42 0 n/a n/a n ,IGSCC 1 TOTAL Made 178 30 159 113 nJa 16 Notes (1) Systems are described in Tables 3.1a and 3.1b.
(2) High Safety Significant (3) Degradation Mechanisms No more than 10% of HSS piping welds are required to be selected for examination. OM selections may be reduced to meet this requirement.
(4) Reactor Coolant Pressure Boundary 1RV (5) For RCPB1FIV (Reactor Coolant Pressure Boundary inside first isolation valve) 2/3 requirement is for total of RPCB and is not required to be met per system.
(6) Reactor Coolant Pressure Boundary outside containment (7) Break Exclusion Region