ML14337A044
| ML14337A044 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 11/19/2014 |
| From: | Berg M South Texas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| G25, NOC-AE-14003194, STI: 33976295 | |
| Download: ML14337A044 (10) | |
Text
WN3Fr Nuclear Operating Company South Teas Proed Electric Genertlng Station PO. Box 289 Wadsworth, Tewas 77483 November 19, 2014 NOC-AE-14003194 File No.: G25 10 CFR 50.55a U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 1 Docket No. STN 50-498 Request for Relief from Code Case N-770-1, Subsection 2400 and Table 1 Inspection Frequency of Reactor Vessel Cold Leg Nozzle to Safe-end Welds.
(Relief Request RR-ENG-3-17)
In accordance with the provisions of 10 CFR 50.55a(a)(3)(ii), STP Nuclear Operating Company (STPNOC) requests relief for South Texas Project (STP) Unit 1 for performing the reactor vessel cold leg nozzle to safe-end weld inspections, covered by ASME Code Case N-770-1, by the currently scheduled outage 1 RE19 (Fall 2015, Unit 1). This relief request proposes extending the inspection period by one operating cycle and performing the Code Case N-770-1 inspections in conjunction with the implementation of an approved stress improvement process to mitigate primary water stress corrosion cracking (PWSCC) in the Hot and Cold leg nozzle to safe-end welds. The purpose of this relief request is to extend the Code Case N-770-1 inspections by one cycle, approximately 18 months, until Refueling Outage 1 RE20 scheduled for Spring 2017.
10CFR50.55a(g)(6)(ii)(F)(1) that became effective July 21, 2011, requires that the STP Inservice Inspection program implement Code Case N-770-1 "Examination Requirements for class 1 piping and nozzle dissimilar-metal butt welds". STPNOC has determined that compliance with code inspection requirements is impractical and would result in unnecessary hardship without a compensating increase in the level of quality and safety.
By performing the cold leg weld inspections in conjunction with an approved stress improvement process during Refueling Outage 1 RE20, STPNOC would reduce unnecessary radiation exposure to personnel, and perform two work evolutions during the same time period. This request is for a one cycle, approximately 18 month, extension scheduled for Spring 2017.
STPNOC requests NRC review and approval of this relief request by May 2015, to support the use of the proposed inspection date extension when authorized, as required by 10 CFR 50.55a(a)(3).
STI: 33976295
NOC-AE-14003194 Page 2 of 3 There are no commitments in this letter.
If there are any questions, please contact Rafael Gonzales at 361-972-4779, or me at 361-972-7030.
Michael Berg Manager Design Engineering/
Testing and Programs rjg
Attachment:
SOUTH TEXAS PROJECT UNIT 1, Request for Relief from Code Case N-770-1, Subsection 2400 and Table 1 Inspection Frequency of Reactor Vessel Cold Leg Nozzle to Safe-end Welds. (Relief Request RR-ENG-3-17)
NOC-AE-14003194 Page 3 of 3 cc:
(paper copy)
(electronic copy)
Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, Texas 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8B1) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 A. H. Gutterman, Esquire Morgan, Lewis & Bockius, LLP John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Cris Eugster L.D. Blaylock City Public Service Peter Nemeth Crain Caton & James, P.C.
C. Mele John Wester City of Austin Robert Free Richard A. Ratliff Texas Department of State Health Services Balwant K. Singal U. S. Nuclear Regulatory Commission
Attachment NOC-AE-14003194 Page 1 of 7 SOUTH TEXAS PROJECT UNIT 1 Request for Relief from Code Case N-770-1, Subsection 2400 and Table 1 Inspection Frequency of Reactor Vessel Cold Leg Nozzle to Safe-end Welds.
(Relief Request RR-ENG-3-17)
A. ASME Component(s) Affected The affected components are STP Unit 1 reactor vessel cold leg nozzle to safe-end welds (Table 1), which are Alloy 600 welds subject to Code Case N-770-1 (Reference 1).
Table 1 - STP Unit 1 reactor vessel cold leg nozzle to safe-end welds UNIT I CATEGORY ITEMNO STP COMP ID COMPDESC
SUMMARY
NO N-770-1 B
101350 RPV1-N2ASE SAFE END TO RPV LOOP A
INLET NOZZLE N-770-1 B
101485 RPV1-N2BSE SAFE END TO RPV LOOP B
INLET NOZZLE N-770-1 B
101635 RPV1-N2CSE SAFE END TO RPV LOOP C
INLET NOZZLE N-770-1 B
101775 RPV1-N2DSE SAFE END TO RPV LOOP D
INLET NOZZLE B. Applicable ASME Code Edition and Addenda ASME Section Xl 2004 Edition (Reference 2)
Code Case N-770-1 as referenced in 10CFR50.55a(g)(6)(ii)(F)(1).
C. Applicable ASME Code Requirement Table 1 of Code Case N-770-1, requires volumetric examination of essentially 100% of Inspection Item B pressure retaining welds once every second inspection period not to exceed 7 years. This is the third In-service Inspection (ISI) inspection interval beginning September 25, 2010 through September 24, 2020.
D. Reason for Relief from Code Requirements Relief is being requested at this time to extend the cold leg weld inspections one cycle (approximately 18 months) to spring 2017 for Refueling Outage 1RE20. STPNOC will be performing mitigation of primary water stress corrosion cracking (PWSCC) in the Cold leg nozzle to safe-end welds using a stress improvement process, and to perform this evolution a critical core barrel lift has to be performed. By receiving relief to move the inspection by one operating cycle, the site can leverage performing the inspection and
Attachment NOC-AE-14003194 Page 2 of 7 the mitigation of PWSCC during the same evolution, thus reducing the risk of a critical lift, and adhering to best "As Low As Reasonably Achievable" (ALARA) practices.
E. Proposed Alternative and Basis for Use:
10CFR50.55a(a)(3) states in part:
"Proposed alternatives to the requirements of paragraph (g), or portions thereof, may be used when authorized by the Director of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. Any proposed alternatives must be submitted and authorized prior to implementation. The applicant or licensee shall demonstrate that:
(i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
STPNOC believes that the proposed alternatives of this request provide an acceptable level of quality and safety. STPNOC proposes a one time extension to Code Case N-770-1, Table 1, Inspection Item B, volumetric examinations from a period not to exceed 7 years to a one time period not to exceed 8 years for STP Unit 1.
During the Unit 1 Spring 2017 outage, STPNOC will be performing the mitigation of PWSCC on the reactor vessel inlet and outlet nozzle to safe end Alloy 82/182 dissimilar metal (DM) welds. STPNOC plans to use a non-welded stress improvement method (meeting the performance criteria of Code Case N-770-1 Appendix 1) as the mitigation process to minimize the potential of PWSCC by permanently eliminating the tensile stress through approximately the inner 50% of the DM weld wall thickness. The NRC and the commercial nuclear industry recognize non-welded stress improvement method as a permanent solution to eliminate the risk of PWSCC. Examination of Code Case N-770-1 Item B (Cold leg) welds are performed from the ID at Unit 1 due to extremely limited access provisions from the outside surface of the pipe. The STP Item A-2 and Item B welds are located inside a "sandbox" which was installed during original plant construction after all welding was completed. The inspection of Item B (Cold leg) welds from the ID requires removal of the reactor vessel core barrel.
The removal of the reactor vessel lower internals assembly (core barrel) is considered to be a critical lift due to the weight of the component, the tight clearances involved, and the radiation emitted by the assembly. For these reasons, only personnel directly involved with the movement of the internals are typically allowed in containment during the evolution.
Remote cameras are utilized to allow most personnel involved with the lift to be outside of the refueling cavity area to minimize personnel radiation exposure. The lower internals lifts are performed solely by viewing cameras. If the need arises the Polar Crane operator is instructed to sit on the floor of the cab or behind shielding and not to raise his head above the cab area of the crane to maintain his radiation dose as low as reasonably achievable (ALARA).
Attachment NOC-AE-14003194 Page 3 of 7 For STP, removing the core barrel requires that it be raised above the refueling cavity water level during transfer from the reactor vessel to the storage stand location. As can be expected, the radiation exposure levels for this activity can be high and necessitate unrelated work to stop, evacuation of personnel from containment, and installation of shielding for the polar crane operator(s). In addition, the dose rates in the area would increase due to the presence of the reactor vessel in the temporary storage location. By aligning the N-770-1 inspection with the non-welded stress improvement method activity, and performing two work evolutions during the same time period, this would reduce unnecessary radiation exposure to personnel. Eliminating the need to remove the core barrel and lower internals during 1 RE1 9 could save approximately 610.5 mrem of dose.
The total dose attributed to removal of the core barrel and lower internals was estimated based on data from 2RE14, the most recent outage when the core barrel was removed. The total dose for the actual work activities to remove and install the reactor core barrel and lower internals during 2RE14 was 123 mrem. The core barrel was transferred to the Lower Internal Storage Area (LISA) where it was stored underwater for 13 days. The dose rates in the vicinity of the LISA with the core barrel present were compared to the dose rates without the core barrel present. The approximate increase in dose rates in the general area walkway was 1.3 mrem/hour (surveys #47960 and 47718). Dose rates were taken on the south end of the 68' elevation of the Reactor Containment Building (RCB) which is a general area walkway and a common travel path for workers inside containment. During the 13 days that the core barrel was stored in this area, workers could have received additional dose due to the higher area dose rates totaling approximately 487.5 mrem (see assumptions below). The total dose associated with moving and storing the core barrel and lower internals is 610.5 mrem.
Assumptions
- 1. The total time the core barrel remained in the LISA, and thus, caused increase dose rates in the general area walkway was 13 days.
- 2. The total RWP-hours during those 13 days was approximately 37,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />.
- 3. The total number of hours that workers may have spent in the vicinity of the 68' with higher dose rates is approximately 1% of the total RWP-hours = 375 hours0.00434 days <br />0.104 hours <br />6.200397e-4 weeks <br />1.426875e-4 months <br />.
- 4. The average increase in dose rates in the general area walkway was 1.3 mrem/hour.
Calculation: 375 hours0.00434 days <br />0.104 hours <br />6.200397e-4 weeks <br />1.426875e-4 months <br /> x 1.3 = 487.5 mrem Operating experience on Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82/182 welds shows that weld repairs performed during original plant construction are a significant contributor in the initiation and propagation of cracking. A review of the construction records and a weld repair search performed for the STP Unit 1 Reactor Vessel nozzle Alloy 82/182 welds did not identify any weld repairs performed on these welds during original plant construction.
Attachment NOC-AE-14003194 Page 4 of 7 In April 2014, ultrasonic (volumetric) and eddy current (surface) exams were performed on the STP Unit 1 Hot Leg welds and no indications were identified. In fall 2015, ultrasonic (volumetric) and eddy current (surface) exams are scheduled to be performed on the STP Unit 1 Cold Leg welds to meet the requirement of N-770-1. The absence of any indications in the Hot Leg welds in 2014 provides added assurance that the one time extension of the inspection of the Cold Leg welds by approximately 18 months provides an acceptable level of quality and safety.
STP will perform non-welded stress improvement method on the reactor vessel inlet and outlet nozzle to safe end welds during the 1 RE20 refueling outage scheduled for spring 2017. This proposed approach reduces radiological exposure and personnel safety hazards associated with critical lifting of the reactor vessel lower internals assembly (core barrel). Therefore, deferral of the Cold Leg Nozzle inspections for STP Unit 1 refueling outage would eliminate the increased radiation exposure associated with the removal of the core barrel.
Technical Basis Electric Power Research Institute (EPRI) Technical Report for Materials Reliability Program:
PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension, MRP-349 (Reference 3) provides the basis for extension of the current volumetric inspection interval for the Reactor Pressure Vessel (RPV) Cold Leg (CL) dissimilar metal welds from every second inspection period or 7 years, as currently required by Code Case N-770-1, to 8 years in the current inspection interval. In summary, the basis for one time extension of Code Case N-770-1, Table 1, Inspection Item B, volumetric examinations from a period of not to exceed 7 years to a period of not to exceed 8 years is:
(1) there has been no service experience with cracking found in RPV CL DM welds, (2) crack growth rates in RPV CL DM welds are small, and (3) likelihood of cracking or through wall leaks is very small in RPV CL DM welds. This technical basis demonstrates that the re-examination interval can be extended to 8 years while maintaining an acceptable level of quality and safety.
Service Experience The STP Unit 1 CL welds were last examined in the fall 2009 using remote mechanized examinations from the Inside Diameter (ID) in accordance with Appendix VIII using performance demonstrated methods where 100% of the flaws (in the test specimens) were detected. In addition, an eddy current examination was performed on the inside (or wetted) surface to interrogate for surface connected flaws. No recordable indications were identified during the 2009 examinations. All volumetric examinations of the STP Unit 1 CL welds previous to 2009 also did not identify any indications requiring resolution. The technique used in site specific exams included 100% coverage for axial and circumferential flaws.
Data is obtained using encoded techniques; therefore, data may be reviewed by multiple qualified examiners. Site specific mock-ups were not used because of the flat, uniform surface associated with performance of these examinations from the ID. These techniques provide a strong assurance that flaws will be detected during inspections. Each STP Unit 1 CL is exposed to approximately 5630F (CL Temperature) during normal plant operation.
Attachment NOC-AE-14003194 Page 5 of 7 Crack Growth Rates (Flaw Tolerance)
All of the flaw tolerance analyses performed to date have shown that the critical crack sizes in large-diameter butt welds operating at CL temperatures are very large. Assuming that a flaw initiates, the time required to grow to through-wall is in excess of 20 years in most cases analyzed. The time to grow from a through-wall leak to a crack equal to the critical crack size can be in excess of 40 years.
More recent analyses have been performed for the RPV nozzles using through-wall residual stress distributions that were developed based on the most recent guidance. These analyses have shown that the flaw tolerance of these locations is high and postulated circumferential flaws will not reach the maximum ASME allowable depth in less than 10 years. Crack growth analysis is given for limiting plants part-circumferential through-wall flaws in Table 5-2 of MRP-349.
Probability of Cracking or Through Wall Leaks Analyses have been performed to calculate the probability of failure for Alloy 82/182 welds using both probabilistic fracture mechanics and statistical methods. Both approaches have shown that the likelihood of cracking or through-wall leaks, in large-diameter CL welds, is very small. Furthermore, sensitivity studies performed using probabilistic fracture mechanics have shown that even for the more limiting high temperature locations, more frequent inspections than required by Section Xl, such as that in MRP-139 or Code Case N-770, have only a small benefit in terms of risk.
Though past service experience may not be an absolute indicator of the likelihood of future cracking, the experience does give an indication of the relative likelihood of cracking in CL temperature locations versus hot leg temperature locations. While there is a significant amount of PWSCC service experience in hot leg locations, the number of indications in large-bore butt welds is still small relative to the number of potential locations. Also, all indications have been detected before they were a safety concern. Therefore, if Hot Leg PWSCC is a leading indicator for CL PWSCC, and the higher frequency of inspections will be maintained for the hot leg locations, it is reasonable to conclude that a moderately less rigorous inspection schedule would be capable of detecting any CL indications before they became large enough to be a concern.
Attachment NOC-AE-14003194 Page 6 of 7 Table 2 below provides a summary of the latest Nozzle to Safe-End Welds inspection for STP Unit 1 (1 RE1 8) and evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the STP Unit 1 Dissimilar Metal Welds.
Table 2:
Information Pertaining to Class 1 Piping and Nozzle Dissimilar-Metal Butt Welds Inspection -STP Unit 1 Inspection During the most recent inservice inspection, all Code Case N-770-1 Methodology:
Inspection Item A-2 (Hotleg) welds, were governed by the ASME Section Xl, 2004 Edition, with no Addenda, Code Case N-770-1 as incorporated by reference 10CFR50.55a.
Number of past Cold Leg examinations were performed with 10-Year inservice inspections:
inspections 1RE08 (1999) and 1RE15 (2009).
Number of There were no recordable indications identified during the most recent indications inservice inspection.
found:
Proposed The third inservice inspection is currently scheduled to be performed inspection in 2015 and 2020. Pending approval of this relief request, the Unit 1 schedule for inspection would be (Baseline Examination after Mitigation) 2017 and balance 2027.
of plant life:
F. Duration of Proposed Alternative This request is applicable to STPNOC's inservice inspection program for the third interval for STP Unit 1 and is not to exceed 18 months to spring 2017 for Refueling Outage 1RE20.
Attachment NOC-AE-14003194 Page 7 of 7 G. References
- 1. Code Case N-770-1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of listed Mitigation ActivitiesSection XI, Division 1.
- 2. ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition No Addenda, American Society of Mechanical Engineers, New York.
- 3. EPRI, Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349), August 2012, (1025852).
H. Precedents Relief from this examination requirement to apply the proposed alternative at the South Texas Project was previously approved by the NRC for the following (with ADAMS Accession No. references):
(1) Indian Point Nuclear Generating Unit No. 2 - Request for Relief Request No. IP2-1SI-RR-14, Code Case N-770-1, Reactor Coolant System Cold Leg Nozzle Weld Inspection Frequency Extention (TAC No. ME6801), dated February 2, 2012 (ML120260090).
(2) Arkansas Nuclear One, Unit No. 1 - Request for Alternative ANO1 -ISI-023 to ASME Code Case N-770-1 Volumetric Examination Frequency Requirements for the Fourth 10-Year Inservice Inspection Interval (TAC No. MF3176), dated October 29, 2014 (ML14282A479).
(3) Joseph M. Farley Nuclear Plant, Units 1 and 2 - Request for Alternative FNP-ISI-13 Regarding Deferral of Inservice Inspection of Reactor Pressure Vessel Cold Leg Nozzle Dissimilar Metal Welds (TAC Nos. ME9739 and ME 9740), dated August 8, 2013 (ML13212A176).