ML23109A070
| ML23109A070 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/04/2023 |
| From: | Cayetano Santos NRC/NRR/VPOB |
| To: | Brown K Southern Nuclear Operating Co |
| Shared Package | |
| ML23109A067 | List: |
| References | |
| Download: ML23109A070 (1) | |
Text
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE REQUIREMENTS VEGP 3&4ISI1ALT18 APPLICATION OF VT1 VISUAL EXAMINATION METHODOLOGY FOR INSERVICE INSPECTION OF REACTOR VESSEL NOZZLE INNER RADIUS SECTIONS SOUTHER NUCLEAR OPERATING COMPANY, INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 DOCKET NOS. 52025 AND 52026
1.0 INTRODUCTION
By letter dated November 4, 2022 (Agencywide Documents and Access Management Systems Accession No. ML22308A157), Southern Nuclear Operating Company, Inc. (the licensee, SNC) requested authorization to use alternatives to the requirements of Section XI, Table IWB25001, Examination Category BD, Item B3.100 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, 2017 Edition at Vogtle Electric Generating Plant (VEGP), Units 3 and 4. The proposed alternative would allow SNC to perform a VT1 visual examination for the inservice inspection of the inner radius sections of the reactor vessel inlet, outlet, and direct vessel injection (DVI) nozzles in lieu of the volumetric examination required by the ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100.
2.0 REGULATORY EVALUATION
Section 50.55a(g)(4)(i) of Title 10 of the Code of Federal Regulations (10 CFR) requires that ASME Code Class 1 components (including their supports) in nuclear power facilities licensed under a combined license under 10 CFR Part 52, must, in the initial 120-month interval, meet inservice inspection requirements of the latest ASME Code Section XI incorporated by reference in 10 CFR 50.55a 18 months prior to planned initial fuel loading. Per 10 CFR 50.55a(z), alternatives to the requirements of 10 CFR 50.55a may be used when authorized by the Director, Office of Nuclear Reactor Regulation. In proposing alternatives, the licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety; or (2) compliance would result in hardship or unusual difficulty without a compensating increase in quality and safety.
Enclosure
3.0 TECHNICAL EVALUATION
3.1 SNCs Proposed Alternative The components affected by this request are the ASME Code Class 1 inlet, outlet, and DVI nozzles in the VEGP Units 3 and 4, reactor vessel upper shells. Each reactor vessel has four 22-inch inner diameter inlet nozzles, two 31-inch inner diameter outlet nozzles, and two 6.81-inch inner diameter DVI nozzles (8-inch schedule 160 pipe connections) designed in accordance with ASME Code,Section III, Subsection NB (16 total nozzles, 8 for Unit 3 and 8 for Unit 4). The nozzles are fabricated of SA508, Grade 3, Class 1 ferritic steel forgings clad on the inner diameter surface. The first cladding layer is Type 309L stainless steel and the subsequent cladding layers are Type 308L stainless steel.
The ASME Code of Record for the initial 120-month interval for inservice inspection of VEGP Units 3 and 4, is the 2017 Edition of ASME Code,Section XI. The inspection requirements for ASME Code Class 1 components are provided in ASME Code,Section XI, Subsection IWB. In accordance with ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100, volumetric (e.g., ultrasonic) examination is the required examination method for the reactor vessel nozzle inner radius section.
SNC proposed to perform a VT1 visual examination for the inservice inspection of the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles for VEGP Units 3 and 4 in lieu of the volumetric examination required by ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100. The proposed VT1 visual examination method will use an underwater camera system attached to a submersible and will be performed in accordance with ASME Code,Section XI. The proposed VT1 visual examination will cover essentially 100 percent of the examination volume as defined in ASME Code,Section XI, Figure IWB25007(b). SNC noted that the service-induced flaw mechanisms (fatigue) for the inner radius sections of the reactor vessel nozzles will be associated with the inner diameter surface of the cladding and further noted that VT1 visual examinations are sufficient to detect such mechanisms well before the nozzle suffers degradation of its structural integrity.
SNC stated that it is their intent to adopt ASME Code,Section XI, Code Case N6482 (N6482), Alternative Requirements for Inner Radius Examinations of Class I Reactor Vessel Nozzles,Section XI, Division I, which allows licensees to perform a VT1 visual examination for inservice inspection of reactor vessel nozzles other than boiling water reactor (BWR) feedwater nozzles and operational BWR control rod drive return line nozzles in lieu of the volumetric examination required by ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100 and has been conditionally approved for use by the U.S. Nuclear Regulatory Commission (NRC) in Regulatory Guide (RG) 1.147, Revision 20, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1. SNC further stated that performing a VT1 visual examination for the inservice inspection of the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles for VEGP Units 3 and 4 aligns with the completed preservice inspection.
A fracture assessment was performed by SNC to determine the maximum initial flaw size that will not grow beyond the allowable flaw size at the end of the evaluation period for the projected 60-year life of the plant. Analysis was done considering Level A/B/Test conditions, which are more limiting than Level C/D/Test. Both linear elastic fracture mechanics (LEFM) and elastic plastic fracture mechanics (EPFM) methods were used. For DVI nozzles, the most limiting case of the nozzles, the flaw length acceptance criteria from the ASME Section XI Table IWB35121 is 0.144 detected in a VT1 examination. From the LEFM analysis, the length of a limiting flaw 2
of depth 0.358 would be 1.14, given a 0.22 cladding thickness. SNC determined that its fracture analysis was conservative because the limiting flaw length would be much longer than what would be acceptable under ASME Section XI.
SNC pointed out that using the EPFM results shows that the reactor vessel nozzle can tolerate a flaw over 3 in depth for a 60-year plant life.
All manufacturing and preservice examination were performed to ASME Sections III, V, and XI with no relevant indications shown.
Comparison to Operating Fleet SNC noted that no fatigue cracking in nozzle inside radius sections have been discovered using either ultrasonic or visual examinations during the operating history of commercial pressurized water reactors (PWR). The licensee concluded that nozzle inside radius sections are not readily susceptible to fatigue cracking.
The licensee also noted that visual examinations are common for critical reactor vessel components through the boiling water reactor vessel internals program for BWRs and materials reliability program for PWRs. VT1, enhanced EVT1, and VT3 are employed for critical reactor components. SNC says that all the above shows the adequacy of VT1 visual examination for the nozzle inner radii for the detection of fatigue cracks.
SNC also noted that the nozzles used at VEGP Units 3 and 4 are from forgings, the same fabrication technique used for all operating plant nozzles. SNC claims that the material properties (such as yield strength, ultimate strength, and fracture toughness) for the AP1000 are the same or better compared to the operating fleet. It states that geometries of the subject nozzles are similar to operating reactors because no welds exist in the nozzle corner region.
The water chemistry for the AP1000 at VEGP Units 3 and 4 will follow the same requirements made by the Electric Power Research Institute that the operating fleet follows. As such, the water pH, boron concentration, conductivity, dissolved hydrogen, and oxygen will all be similar to that of the operating fleet. SNC says that a lack of oxygen in the water chemistry precludes corrosion in the carbon steel during operating conditions, which represents about 90 percent of plant lifetime. The stainless steel cladding layered over the carbon steel mitigates the potential for corrosion during shutdown conditions.
During preservice inspection (PSI), VT1 visual examination and liquid penetrant (PT) surface examination were performed on the nozzle inside radius sections of the two outlet nozzles, the four inlet nozzles, and the two DVI nozzles. SNC submitted this proposed alternative to match inservice inspection with preservice inspection.
Based on the above, SNC stated that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).
Duration of Proposed Alternative SNC requests that the alternative applies to the First In Service Inspection (ISI) Interval for VEGP Units 3 and 4.
3.2 Staff Evaluation The reactor vessel inlet, outlet, and DVI nozzles are classified as ASME Code Class 1; therefore, the requirements of ASME Code,Section XI, Subsection IWB, must be applied.
ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100 requires 3
that the reactor vessel nozzle inner radius sections be volumetrically examined. Mandatory Appendix I of ASME Code,Section XI, requires that the procedures, equipment, and personnel used for UT examination of the nozzle inside radius section be qualified by performance demonstration in accordance with ASME Code,Section XI, mandatory Appendix VIII, Supplement 5 or 7. Appendix I also requires that the nozzle inside radius section be examined in two opposing circumferential directions.
To satisfy the inservice inspection requirements, SNC proposed to perform a VT1 visual examination of the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles for VEGP Units 3 and 4 in lieu of the volumetric examination required by ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100. This would align with the preservice examination performed and approved in VEGP 3&4PSI-ALT07 by the staffs safety evaluation issued on September 25, 2018 (ML18263A215 and ML18263A219). SNC intends on performing VT1 visual examinations for the inservice inspections of the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles in accordance with ASME Code,Section XI, Code Case N6482, approved with conditions in RG 1.147.
ASME Code,Section XI, Code Case N6482 allows licensees to perform a VT1 visual examination for inservice inspection of reactor vessel nozzles other than BWR feedwater nozzles and operational BWR control rod drive return line nozzles in lieu of the volumetric examination required by ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100. Code Case N6482 applies only to inservice inspection and is conditionally accepted in RG 1.147, Revision 20. The condition is that the allowable flaw length criteria of ASME Code,Section XI, Table IWB35121, with limiting assumptions on the flaw aspect ratio be used.
The technical basis for Code Case N6482 included a flaw tolerance evaluation using LEFM that resulted in allowable flaw sizes at the end of evaluation period larger than 3 inches for the nozzle inner radius sections of the operating fleet (PWRs and BWRs).
SNC noted that its proposed alternative aligns with the Code Case N6482 requirements. In the rule, Approval of American Society of Mechanical Engineers Code Cases, published on August 16, 2018 (83 FR 40685), the NRC conditioned use of Code Case N6482. The condition is that Code Case N6482 shall not be used to eliminate the volumetric preservice inspection (PSI) examinations and shall not be used to eliminate the preservice or inservice volumetric examination of plants with a combined operating license pursuant to 10 CFR Part 52, or a plant that receives its operating license after October 22, 2015. The statement of considerations for the rule state that the required preservice volumetric examination provides a baseline comparison for future volumetric examination. Elimination of inservice volumetric examinations can be justified in the existing fleet because of good operating experience, showing no inner radius cracking in the relevant nozzles. However, new reactor designs have no inspection history or operating experience available to support eliminating volumetric examination of the nozzles in question. NRC conditions on a code case do not preclude a licensee from submitting an alternative similar to the conditioned code case. As previously discussed, in accordance with 10 CFR 50.55a(z), the licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety; or (2) compliance would result in hardship or unusual difficulty without a compensating increase in quality and safety. As part of its site-specific review, the staff evaluated the applicability of the operating experience for similarly designed reactor vessel nozzles for the current operating PWR fleet to the VEGP Units 3 and 4 reactor vessel inlet, outlet, and DVI nozzles. Specifically, the staff reviewed design, materials, water chemistry, stresses, fatigue and flaw tolerance evaluations, and the proposed VT1 visual examination.
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Design, Materials, and Water Chemistry VEGP Units 3 and 4, reactor vessel inlet and outlet nozzles are of similar size and geometry of similarly designed reactor vessel inlet and outlet nozzles of the current operating PWR fleet.
However, VEGP Units 3 and 4, reactor vessel DVI nozzles differ in geometry and size compared with the reactor vessel inlet and outlet nozzles for both VEGP Units 3 and 4, and the current operating PWR fleet. The inner diameter of subject DVI nozzles (6.81 inches) is smaller and includes a 4-inch venturi for the purpose of limiting high-pressure blowdown flow. In addition, the inner radius section thickness of reactor vessel DVI nozzles is thinner than that of the current operating PWR fleet. Like the current operating PWR fleet, subject reactor vessel inlet, outlet, and DVI nozzles do not have a weld in the inner radius section of the nozzles.
Similar to the current operating PWR fleet, which has shown no cracking in the inner radius of the vessel inlet, outlet, and DVI nozzles, the subject reactor nozzles are fabricated of SA508, Grade 3, Class 1, ferritic steel forgings clad on the inner diameter surface with multiple stainless steel cladding layers.
The licensee stated that the water chemistry that VEGP Units 3 and 4 reactor vessel inlet, outlet, and DVI nozzles will be exposed to follows the same Electric Power Research Institutes water chemistry guidelines as the current operating PWR fleet. The licensee further stated that in comparison to some operating PWRs, the water chemistry may be better for VEGP Units 3 and 4 due to following a more recent version of the industry guidelines.
Stresses and Fatigue and Flaw Tolerance Evaluations SNC stated that the pressure stresses in the AP1000 nozzle inner radius sections are similar to operating vessels of a similar size and geometry; however, the stresses from the thermal transients differ for the AP1000 design because it includes more transients overall due to the passive cooling design of the plant and has more transient cycles due to it being designed for a 60-year life. Given that the stresses from the thermal transients are different from the operating vessels, the staff also reviewed the fatigue and flaw tolerance evaluations for the AP1000 reactor vessel inlet, outlet, and DVI nozzles.
SNC used the stresses from a finite element analysis model to perform flaw tolerance evaluations of the subject nozzles. SNC determined the allowable end of evaluation period flaw sizes (depth within the underlying base metal) for the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles for the AP1000 plant design using LEFM for Level A/B/Test conditions and Level C/D conditions in accordance with ASME Code,Section XI, Subsection IWB3600; and using EPFM for only Level A/B/Test conditions (more limiting than Level C/D conditions) in accordance with ASME Code,Section XI, Code Case N749, Alternative Acceptance Criteria for Flaws in Ferritic Steel Components Operating in the Upper Shelf Temperature Range, as conditioned in 10 CFR 50.55a. SNC stated that the EPFM provides a more realistic fracture assessment considering the resistance to crack extension of the ductile nozzle material. The LEFM for Level A/B/Test conditions resulted in average allowable end of evaluation period flaw depth for the DVI nozzles of 0.36, which corresponds to a flaw length of 1.14, based on a flow depth to length ratio of 0.5 and a cladding thickness of 0.22. The staff finds this acceptable because the limiting flaw is larger than the ASME acceptance criteria of a flaw length of 0.144. The LEFM flaw tolerance results for the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles for the AP1000 plant design showed tolerance for smaller flaws (1.14) than the LEFM flaw tolerance results for the operating fleet (greater than 3). However, the staff recognizes that the flaw tolerance results for the AP1000 reactor vessel inlet, outlet, and DVI nozzles were calculated for 60 years, and, therefore, would be closer to the flaw tolerance results for the operating fleet if calculated for 40 years like the operating fleet.
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The staff noted that the inner radius thicknesses for the AP1000 inlet and outlet nozzles and operating fleet inlet and outlet nozzles are similar, and therefore would have similar flaw sizes.
Because the inner radius thickness for the DVI nozzles is thinner than the operating fleet, the corresponding flaw depth that can be tolerated would be smaller for the AP1000 plant but would have similar flaw depth to inner radius thickness ratio as the operating fleet.
Proposed VT1 Visual Examination SNC proposed to perform a VT1 visual examination for the inservice inspection of the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles. The licensee stated that the proposed VT1 visual examination method will use an underwater camera system attached to a submersible and will be performed in accordance with ASME Code,Section XI. The proposed VT1 visual examination will cover essentially 100 percent of the examination volume as defined in ASME Code,Section XI, Figure IWB25007(b). In addition, SNC stated that all stress analysis locations, including the limiting stress cuts in the nozzle inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles, evaluated for fatigue and flaw tolerance, will be examined during the preservice inspection using the proposed VT1 visual examination.
SNC committed to apply the allowable flaw length criteria of ASME Code,Section XI, Table IWB35121, with a/l limited to 0.5 and a/t limited to 2.5 percent, as the VT1 visual examination acceptance standard (NRC condition on use of Code Case N6482) where a is the flaw depth and l is the flaw length. Applying this allowable flaw length criteria, SNC determined the VT1 visual examination acceptance standard for the reactor vessel inlet, outlet, and DVI nozzles as 0.392, 0.384, and 0.144, respectively. A VT1 visual examination finding exceeding the acceptance standards would result in repair/replacement, reexamination, and regulatory review in accordance with ASME Code,Section XI, Subsections IWB3113 and IWB3114, to ensure fitness for service. Based on the flaw tolerance evaluations, the VT1 visual examination acceptance standards for the reactor vessel inlet, outlet, and DVI nozzles are smaller than the maximum initial flaw lengths for the reactor vessel inlet, outlet, and DVI nozzles, even for the conservative LEFM. The staff finds the use of Code Case N6482 acceptable because the acceptable flaw length is smaller than the maximum initial flaw lengths determined using LEFM for the reactor vessel inlet, outlet, and DVI nozzles. Any flaw exceeding the acceptance criteria will be repaired/replaced, reexamined, and reviewed in accordance with ASME Code,Section XI, Subsections IWB3113 and IWB3114 to ensure fitness for service.
Performing the VT1 visual examination in accordance with ASME Code,Section XI, demonstrates that SNCs examination methodology (i.e., personnel and equipment) has a resolution sensitivity capable of detecting flaws smaller than the maximum initial flaw lengths for the reactor vessel inlet, outlet, and DVI nozzles, even for the more conservative LEFM; and flaws with lengths smaller than the acceptance standards for the reactor vessel inlet, outlet, and DVI nozzles.
Based on its review, the staff finds that SNC has demonstrated that the proposed alternative provides an acceptable level of quality and safety because the VEGP Units 3 and 4 reactor vessel inlet, outlet, and DVI nozzles:
Are fabricated in the same manner as the operating fleet, including no weld in the inner radius section, and with the same materials.
Will be exposed to water chemistry that is the same or better than the operating fleet.
Demonstrate tolerance for flaws within an acceptable margin to the flaw tolerance for the operating fleet.
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Will be inspected with a VT1 visual examination performed in accordance with ASME Code,Section XI, as conditioned by 10 CFR 50.55a, that will cover essentially 100 percent of the examination volume as defined in ASME Code,Section XI, Figure IWB25007(b). The examination volume includes the limiting stress cut locations for the inner radius sections of the reactor vessel inlet, outlet, and DVI nozzles.
Will be inspected with a VT1 visual examination method capable of detecting the type and sizes of flaws expected.
4.0 CONCLUSION
As set forth above, the staff determines that the proposed use of Code Case N6482 as an alternative to the requirements of the 2017 Edition of ASME Code,Section XI, Table IWB25001, Examination Category BD, Item B3.100 provides an acceptable level of quality and safety. Accordingly, staff concludes that SNC has adequately addressed all of the regulatory requirements in 10 CFR 50.55a(z)(1). The staff also recognizes that this alternative request aligns inservice inspections with previously performed preservice inspections of these nozzles. Therefore, the staff authorizes VEGP 3&4ISI1ALT18 for the inservice inspection during the first inservice inspection interval at VEGP Units 3 and 4.
All other requirements of ASME Code,Section XI, and 10 CFR 50.55a for which an alternative has not been specifically requested and authorized, remain applicable.
5.0 REFERENCES
- 1. VEGP 3&4-ISI1-ALT-18, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1):
Alternative for Use of Code Case N-648-2 for lnservice Inspection of the Reactor Vessel Nozzle Inside Radius Sections, dated November 4, 2022 (ML22308A158).
- 2. VEGP 3&4PSI-ALT07, Request for Alternative: Application of VT-1 Visual Examination Methodology for Preservice Inspection of the Reactor Vessel Nozzle Inner Radius Sections, dated July 6, 2017 (ML17192A125).
- 3. Vogtle Electric Generating Plant, Units 3 and 4 - Request for Alternative: Application of VT-1 Visual Examination Methodology for Preservice Inspection of Reactor Vessel Nozzle Inner Radius Sections (VEGP 3&4-PSI-ALT-07) (EPID L-2017-LLA-0085), dated September 25, 2018 (ML18263A213).