ML121170481
| ML121170481 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/24/2012 |
| From: | Mogren T Ginna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML121170481 (10) | |
Text
WPLNRC-1002545 Thomas Mogren RE. Ginna Nuclear Power Plant, LLC Manager - Engineering Services 1503 Lake Road Ontario, New York 14519-9364 585.771.5208 585.771.3392 Fax C ENG Thomas.Mogren@cengilc.com a joint venture of Conselto *-#eDF April 24, 2012 U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTENTION: Document Control Desk
SUBJECT:
R.E. Ginna Nuclear Power Plant Docket No. 50-244 Fifth Interval Inservice Inspection Program Submittal of 10 CFR 50.55a Relief Request Number ISI-06
REFERENCE:
(1) Letter from T. Mogren, Ginna LLC, to NRC Document Control Desk, dated March 16, 2012,
Subject:
ASME Code Section XI Evaluation of the Bottom Mounted Instrumentation (BMI) Penetration Nozzle A86 at the R.E. Ginna Nuclear Power Plant Pursuant to 10 CFR 50.55a(a)(3)(ii), R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) hereby requests NRC approval of the enclosed request for the Fifth Interval Inservice Inspection Program.
During the 2011 refueling outage, all of the Bottom Mounted Instrumentation Nozzles (BMI) were volumetrically inspected in accordance with Ginna's Fifth Interval 10-Year, Inservice Inspection (ISI)
Plan and Fourth Interval ISI Relief Request Number 24. Except for BMI Penetration Nozzle A86, no indications were found. Two indications were found in Nozzle A86 and, by Reference (1), Ginna LLC submitted an analytical evaluation for indication number 1. Indication number 2 was acceptable. It was concluded in Reference (1) that the two indications were not the result of Primary Water Stress Corrosion Cracking (PWSCC) and that they were introduced during the original fabrication of the vessel. ASME Section XI Code states that the area containing the flaw shall be subsequently reexamined during the next three inspection periods listed in the schedule of the inspection program of IWB-2400. This reexamination schedule creates an additional risk management burden and will result in more radiation worker exposure, without a compensating increase in quality or safety.
Relief Request Number ISI-06 (Enclosure 1) is being submitted for the fifth 10-year ISI interval to request a proposed alternative to complying with the code reexamination requirement. Enclosure 1 describes how the proposed alternative provides an acceptable level of quality and safety. Approval is requested by October 31, 2012.
A-V 7
Document Control Desk April 24, 2012 Page 2 There are no new commitments identified in this correspondence. Should you have any questions regarding this request, please contact Mr. Thomas Harding at (585) 771-5219 or Thomas. HardinaJarDcenqllc.com.
,_Xery truly yours, Thomas Mogren cc: W.M. Dean, NRC M.C. Thadani, NRC Ginna Resident Inspector, NRC : Relief Request Number ISI-06 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii) Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections
ENCLOSURE I Relief Request Number ISI-06 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections
Relief Request Number ISI-06 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections
- 1. ASME Code Component(s) Affected Class 1, Reactor Pressure Vessel, Bottom Mounted Instrumentation (BMI) Penetrations. This Relief Request is applicable to BMI Penetration Nozzle number A86.
- 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code Section Xl, 2004 Edition, No Addenda.
- 3. Applicable Code Requirement Paragraph IWB-2420 (b) which states that if a component is accepted for continued service in accordance with IWB-3132.3 or IWB-3142.4, the areas containing flaws or relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the inspection program of IWB-2400.
- 4. Reason for Request Relief Request Number ISI-06 is being submitted to provide an alternative to ASME Code Section Xl Paragraph IWB-2420(b) of performing successive ultrasonic (UT) examinations each period on Reactor Pressure Vessel BMI Penetration Nozzle number A86. In accordance with 10CFR50.55a(a)(3)(ii) compliance with the specific requirements of IWB-2420(b) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Alternatively, it is proposed to perform one successive exam, two periods following the initial detection of an original fabrication flaw in BMI Penetration Nozzle A86 and maintain detailed visual examination for leakage every refueling outage. The specific background and information regarding this relief request are as follows.
The Nuclear Regulatory Commission (NRC) has included, effective September 8, 2008, Code Case N-722 with conditions in IOCFR50.55a(g)(6)(ii)(E)(2) through (4). These new requirements included the need for a bare metal visual examination of the Reactor Pressure Vessel BMI Penetration Nozzles during every other refueling outage. The lower bottom exterior portion of the R. E. Ginna Reactor Pressure Vessel has a bitumastic paint with a zinc filler that has been there since original operation. This paint precludes a bare metal visual exam. Due to the hardship that would be imposed to remove the paint, R. E. Ginna Nuclear Power Plant Page 1 of 7
Relief Request Number ISI-06 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections submitted Relief Request Number 24 [4] on May 22, 2009. This Relief Request included, among other alternatives, the alternative requirement to perform an UT examination of all 36 Reactor Pressure Vessel BMI Penetration Nozzles.
All of the BMI Penetration Nozzles were UT examined in accordance with ISI Relief Request Number 24 during the spring 2011 refueling outage. All 36 of the BMI Penetration Nozzles were examined by time-of-flight diffraction (TOFD) in both axial and circumferential oriented scan directions, 0 degree ultrasonic, 45 degree shear wave ultrasonic, and also an eddy current non-destructive examination method. BMI Penetration Nozzle A86 was the only nozzle that exhibited ultrasonic indications requiring disposition. Penetration Nozzle A86 had two (2) ultrasonic indications. No BMI Penetration Nozzles exhibited any eddy current examination indications.
Indication number 1 in Penetration Nozzle A86 was seen with the time-of-flight-diffraction ultrasonic examination and was classified as a circumferentially oriented planar flaw which extends into the nozzle wall from the weld-to-nozzle interface. The dimension of indication number I was determined to be 1.35 inch circumferentially by 0.161 inch in through-wall depth. The nozzle wall thickness is approximately 0.59 inches, making the indication approximately 27 percent of the nozzle wall in depth. The indication was not seen with either the 0 degree or 45 degree shear wave ultrasonic exam or the eddy current examination.
With the assistance of the Electric Power Research Institute (EPRI) NDE Center, the time-of-flight ultrasonic data for indication number I was compared to similar data from i-groove welds at other plants. It was determined that indication number 1 has many signal characteristics similar to indications found in retired Reactor Vessel Head Control Rod Drive penetrations at the nozzle-to-weld interface. Those indications were destructively verified as being created by fabrication, most likely as a result of an in-process grind and repair evolution. Indication number I also lacked many of the characteristics of actual Primary Water Stress Corrosion Cracking (PWSCC) indications as outlined in EPRI Report IR-2011-476. This independent review of the data concluded the characteristics of indication number I were more consistent with the response from a metallurgical interface associated with a grind and weld evolution during original fabrication.
Indication number 2 was classified as a laminar flaw, most likely resulting from lack of fusion at the nozzle-to-weld interface. The determined dimension of indication number 2 was 0.16 inch axial by 0.25 inch circumferential by 0.0 inch through-wall depth. This indication was seen in Page 2 of 7
Relief Request Number ISI-06 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections the 0 degree and 45 degree UT examination data. Indication number 2 is an acceptable size not requiring further analytical evaluation [5].
Neither of the Ginna BMI Penetration Nozzle A86 flaws are open to a wetted surface.
Enhanced Visual Examination (EVT-1) was performed on the vessel interior wetted surface and eddy current was performed on the nozzle inside surface of BMI Penetration Nozzle A86. No indication of cracking, pores, or inclusions that could be associated with a wetting path was observed at the weld, nozzle and adjacent vessel surfaces. The EVT-1 examination method, using a 0.044 inch character card, is capable of detecting Stress Corrosion Cracking flaws, as detailed within Pacific Northwest National Laboratory, NUREG/CR-6943, PNNL-16472, "A Study of Remote Visual Methods to Detect Cracking in Reactor Components," October 2007, Prepared for the Division of Fuel, Engineering, and Radiological Research Office of Nuclear Regulatory Research.
A finite element residual stress analysis also shows that the residual axial stresses are low or compressive in the area of the BMI Penetration Nozzle A-86 indications. The propensity for circumferential cracking is low and is less likely than axial cracking because hoop stress in this location exceeds the axial stress. Circumferential PWSCC initiation, which requires contact with the reactor coolant system fluid, is unlikely even without considering the additional beneficial effect of post-weld heat treatment that was performed on Ginna's vessel. The operating temperature of the lower vessel head region is approximately 528 degrees F [6], which translates to a lower probability of initiating PWSCC cracking. The indications are embedded and significantly away from the wetted surface. Therefore, it is concluded that indication number I and 2 in BMI Penetration Nozzle A86 are not the result of PWSCC. It is also concluded that both indications were introduced during the original fabrication of the vessel and have not changed since initial operation.
Although it was concluded that the UT indications were not flaws resulting from an inservice degradation mechanism, under ASME Section Xl Code, 2004 Edition, No Addenda; Paragraph IWB-3132.3 an "Acceptance by Analytical Evaluation" was performed to address indication number 1 [7]. The analytical evaluation was submitted [8] to the NRC as required by paragraph IWB-3144 (b). The Westinghouse analytical evaluation concluded that since the primary axial loads due to internal pressure and other external primary mechanical loading are small for the BMI Penetration Nozzle A86, the maximum end-of-evaluation period allowable flaw depth is 0.446 inch, or 75% of the wall thickness. Fatigue crack growth is the only credible crack growth mechanism for the detected flaw and the resulting fatigue crack growth has been shown to be very small. The Predicted Flaw Depth after 10 additional years of operation is 0.167 inch (an Page 3 of 7
Relief Request Number ISI-06 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections increase of only 0.006 inches). After 20 years of additional operation the analytical depth is computed to be 0.174 inch (an additional 0.007 inch growth). In 40 additional years of operation the evaluated depth is 0.188 inch, far less than the allowed 0.446 inch depth.
ASME Section Xl Code, 2004 Edition, No Addenda; Paragraph IWB-3132.3 "Acceptance by Analytical Evaluation", states that the area containing the flaw shall be subsequently reexamined in accordance with IWB-2420 (b) and (c). Paragraph IWB-2420 (b) states that if a component is accepted for continued service in accordance with IWB-3132.3 or IWB-3142.4, the areas containing flaws or relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the inspection program of IWB-2400. This relief request is to defer the first successive examination from the first period after the initial examination to the second period after the initial examination. This would allow the volumetric examinations on BMI Penetration Nozzle A86 to coincide with a planned removal of the Reactor Vessel Internals at the end of the Fifth 10-year ISI Interval. The ISI and Reactor Vessel Internals Aging Management Program requires Reactor Pressure Vessel (RPV) disassembly to the extent needed to perform the BMI penetration nozzle UT examination in the same manner as initially detected. At that time, the Upper Internals Package, Fuel, and the Core Barrel are removed, providing full access to the BMI penetration nozzles from the reactor vessel interior.
Since 2003, an as-found detailed outer surface visual examination on all 36 BMI penetration nozzles has been performed every refueling outage (about every 18 months) and no evidence of reactor coolant leakage has been observed. Recent detailed visual examinations were performed during the spring 2011 refueling outage, which showed no evidence of reactor coolant leakage on the 36 BMI Penetration Nozzles. These visual examinations have been demonstrated to be effective at detecting small leakage under similar painted conditions in laboratory tests. Detailed visual examinations will continue to be performed each refueling outage throughout the Fifth 10-Year ISI Interval [11].
All BMI Penetration Nozzles were UT examined during the 2011 Refueling Outage (RFO) with the reactor vessel fuel removed and the core barrel and associated lower internals removed.
This degree of reactor vessel disassembly is only undertaken when reactor vessel weld and internals exams are required approximately once per 10 years. Placing the unit into that configuration places an additional risk management burden on the station and will result in more radiation worker exposure. It should also be noted that core barrel removal operation is considered a high risk evolution. No readily available device currently exists that is capable of performing the UT exam without removal of the core barrel. If the examination were to be Page 4 of 7
Relief Request Number ISI-06 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections attempted with a remote device through an in place core barrel and associated lower internals, there would still be the potential of damage to vessel internal items as well as associated difficulty inherent with the location, installation and removal of first of a kind examination equipment. The ultrasonic delivery system would need to travel from the vessel flange, through a small diameter hole in the fuel support plate, and then through the lower internal guides. Gaps would have to be crossed, which could make it difficult to align a UT transducer probe to enter BMI Penetration Nozzle A86, which is located at the lower bottom of the vessel.
This approach would also increase the chance of foreign material becoming lodged in the lower core support area, making removal in this configuration extremely difficult or impossible without further vessel disassembly and removal operations.
- 5. Proposed Alternative and Basis for Use R. E. Ginna Nuclear Power Plant proposes to perform a subsequent UT examination on BMI Penetration Nozzle number A86 once during the 3 rd Period of the Fifth Interval ISI Program.
Indication number I on BMI Penetration Nozzle A86 is very slow growing, and performing the successive UT examination after a short period of time during the 2nd Period of the Fifth Interval ISI Program would do little for trending crack growth. Workers would be exposed to unnecessary dose or contamination. In addition, there would be the possibility of damage to the vessel internals from the UT transducer probe delivery system, or during disassembly of the reactor internals. The indications are not exposed to the primary water environment and therefore not subjected to PWSCC. A detailed visual exam of the outer surface of the lower reactor vessel head will continue to be performed each refueling outage in accordance with R.
E. Ginna Nuclear Power Plant Fifth Interval ISI Program Relief Request IS1-05 [11].
Through the Westinghouse analytical evaluation and continued visual examinations of all BMI Penetration Nozzles every refueling outage, as defined within R. E. Ginna Nuclear Power Plant Fifth Interval ISI Program Relief Request ISI-05, and the performance of a subsequent UT examination of BMI Penetration Nozzle number A86 once during the 3 rd Period of the Fifth Interval ISI Program, reasonable assurance of structural integrity is maintained.
- 6. Duration of Proposed Alternative The duration of this relief request is the remainder of the Fifth Interval ISI Program. The need for further volumetric examination on BMI Penetration Nozzle number A86 will be determined by comparison of the successive 3 rd Period Fifth Interval examination results with the initial Page 5 of 7
Relief Request Number ISI-06 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections examination and the projected results of the Analytical Evaluation. This action will provide reasonable assurance of structural integrity and maintain public and plant safety.
- 7. Precedents Nuclear Regulatory Commission to Rochester Gas & Electric Corporation Safety Evaluation Letter, "Ginna Flaw Indication in the Reactor Vessel Inlet Nozzle Weld - 1989 Reactor Vessel Examination (TAC No. 71906)", Dated July 7, 1989.
- 8. References
- 2. ASME Code Case N-722, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials.
- 3. ASME Section Xl Code, 2004 Edition, No Addenda
- 4. Nuclear Regulatory Commission Safety Evaluation (ML100290926), Dated March 8, 2010. Relief Request Number 24 - Fourth Interval Inservice Inspection Program Proposed Alternative for Bottom Mounted Instrumentation Exams - R. E. Ginna Nuclear Power Plant (TAC NO. ME1364)
- 5. Westinghouse WCAP-17410-P, Revision 0, dated April 2011, "Structural Integrity Evaluation of Reactor Vessel BMI Penetrations to Support Continued Operation: Ginna,"
Westinghouse Proprietary Class 2.
Power Capability Working Group (PCWG) Parameters for Ginna Station Uprate Program, Reactor Coolant Temperature.
- 7. Westinghouse Non-Proprietary Class 3 Report, LTR-PAFM-11-69, Revision 0, Structural Integrity Evaluation of Circumferential Indication in Ginna BMI Nozzle No. A86, July 2011.
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. Relief Request Number ISI-06 R. E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(ii)
Bottom Mounted Instrumentation Nozzle A86 Examination-Successive Inspections
- 8. Letter From Tom Mogren, R. E. Ginna Nuclear Power Plant LLC, to NRC, dated March 16, 2012,
Subject:
ASME Code Section Xl Evaluation on the Bottom Mounted Instrumentation (BMI) Penetration Nozzle A86 at the R. E. Ginna Nuclear Power Plant.
(ML12080A141 and ML12080A142)
- 9. Nuclear Regulatory Commission Safety Evaluation, Dated July 31,2009.
Subject:
R.E.
Ginna Nuclear Power Plant: Safety Evaluation for Relief Request No. 18, Reactor Vessel Weld Examination Extension (TAC NO. MD 9962)
- 10. Pacific Northwest National Laboratory, NUREG/CR-6943, PNNL-16472, A Study of Remote Visual Methods to Detect Cracking in Reactor Components, October 2007, Prepared for Division of Fuel, Engineering and Radiological Research Office of Nuclear Regulatory Research.
- 11. Letter from Tom Mogren, R. E. Ginna Nuclear Power Plant LLC, to NRC, dated December 16, 2011,
Subject:
Relief Request Number ISI-05, Fifth Interval Inservice Inspection Program Proposed Alternative for Bottom Mounted Instrument Examinations.
(MLi1363A074)
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