ML18334A236
| ML18334A236 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 12/21/2018 |
| From: | James Danna Plant Licensing Branch 1 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| Marhsall M, NRR/DORL/LPL, 415-2871 | |
| References | |
| EPID L-2018-LLR-0087 | |
| Download: ML18334A236 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Bryan C. Hanson Senior Vice President December 21, 2018 Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - ALTERNATIVE TO THE REQUIREMENTS OF THE ASME CODE (EPID L-2018-LLR-:0087)
Dear Mr. Hanson:
By letter dated June 8, 2018 (Agencywide Documents Access and Management System Accession No. ML18159A059), Exelon Generation Company, LLC (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of an alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),Section XI requirements at the Nine Mile Point Nuclear Station (Nine Mile Point),
Units 1 and 2, during the fifth and fourth inservice inspection (ISi) intervals, respectively. The purpose of this letter is to provide the results of the NRC staff's review of Relief Requests 15R-05 and 14R-05. The NRC staff will provide separate correspondence regarding the other relief requests.
Pursuant to Title 10 of the Code of Federal Regulations ( 1 O CFR) Section 50.55a(z)( 1 ), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. The proposed alternative would allow the licensee to eliminate the examination of threads in the reactor pressure vessel flange required by Examination Category B-G-1, Item No. 86.40, in Section XI of the ASME Code.
Since the flaw tolerance analysis results meet the ASME Code,Section XI acceptance criterion, the NRC staff determined that the licensee's proposed alternative to the examination requirements for the stud-hole threads in reactor pressure vessel flange of the reactor pressure vessel for both Nine Mile Point, Units 1 and 2, provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1 ). Therefore, the NRC staff authorizes the request for alternative proposed in the licensee's submittal for the fifth 10-year ISi interval for Nine Mile Point, Unit 1 and the fourth 10-year ISi interval for Nine Mile Point, Unit 2.
All other requirements of Section XI of the ASME Code for which relief was not specifically requested and approved in the subject relief request remains applicable, including third party review by the Authorized Nuclear lnservice Inspector.
If you have any questions, please contact the Nine Mile Point Project Manager, Michael Marshall, at (301) 415-2871 or Michael.Marshall@nrc.gov.
Docket Nos. 50-220 and 50-41 O
Enclosure:
Safety Evaluation cc: Listserv Sincerely, Ja es G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUESTS 15R-05 AND 14R-05 NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-220 AND 50-410
1.0 INTRODUCTION
By letter dated June 8, 2018 (Agencywide Documents Access and Management (ADAMS)
Accession No. ML18159A059), Exelon Generation Company, LLC (Exelon or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of an alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),Section XI requirements at the Nine Mile Point Nuclear Station (Nine Mile Point), Units 1 and 2, during the fifth and fourth inservice inspection (ISi) intervals, respectively.
Pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR) 50.55a(z)( 1 ), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. The proposed alternative would allow the licensee to eliminate the examination of threads in the reactor pressure vessel (RPV) flange required by Examination Category B-G-1, Item No. B6.40, in Section XI of the ASME Code.
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.55a(g)( 4) state, in part, that ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in Section XI of the applicable editions and addenda of the ASME Code to the extent practical within the limitations of design, geometry, and materials of construction of the components. The threads in the RPV flange are categorized as ASME Code Class 1 components. Therefore, per 10 CFR 50.55a(g)(4), ISi of these threads must be performed in accordance with Section XI of the applicable edition and addenda of the ASME Code.
The regulations in 1 O CFR 50.55a(z) state, in part, that alternatives to the requirements in paragraphs (b) through (h) of 1 O CFR 50.55a may be authorized by the NRC if the licensee demonstrates that: ( 1) the proposed alternative provides an acceptable level of quality and Enclosure safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request the use of an alternative, and the NRC to authorize the proposed alternative.
3.0
3.1 TECHNICAL EVALUATION
ASME Code Components Affected The affected components are the threads in the RPV flange subject to Examination Category B-G-1, Item No. B6.40, from IWB-2500, Table IWB-2500-1 of the ASME Code,Section XI.
3.2 Applicable ASME Code Edition For the fifth and fourth 10-year ISi intervals at Nine Mile Point, Units 1 and 2, respectively, the Code of record is the 2013 Edition of the ASME Code,Section XI.
3.3 ASME Code Requirements for Which Relief is Requested The licensee has requested an alternative to the examination requirements in Examination Category B-G-1, Item No. B6.40, which is listed in Table IWB-2500-1 of the ASME Code,Section XI. This item requires the licensee to perform, every ISi interval, a volumetric examination of all threads in RPV flange stud holes as shown in Figure IWB-2500-12, "Closure Stud and Threads in Flange Stud Hole," of the ASME Code,Section XI.
3.4 Licensee's Proposed Alternative The licensee is proposing to eliminate the examination of threads in RPV flanges, as required by Examination Category B-G-1, Item No. B6.40, of the ASME Code,Section XI, for the duration of the fifth 10-year ISi interval for Unit 1, and the fourth 10-year ISi interval for Unit 2, or until the NRC approves an applicable alternative in Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1," or other document.
3.5 Licensee's Basis for Use The licensee's request is based on an evaluation by the Electric Power Research Institute (EPRI) documented in EPRI Technical Report No. 3002007626 (EPRI report}, "Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements," dated March 2016 (ADAMS Accession No. ML16221A068). The licensee's submittal included information from the EPRI report regarding the generic stress analysis and the flaw tolerance evaluation, with additional plant-specific information to demonstrate applicability of the EPRI results to Nine Mile Point, Units 1 and 2. The submittal also included information from the EPRI report regarding operating experience and potential degradation mechanisms for threads in RPV flanges. Additionally, the licensee described maintenance activities it performs each time the RPV closure head is removed to detect and mitigate general degradation prior to returning the reactor to service. Specifically, the licensee stated that the threads in the RPV flange are inspected for damage, cleaned, and lubricated prior to reinstallation of the RPV studs.
3.6 Duration of Proposed Alternative The licensee stated that relief is requested for the fifth and fourth ISi intervals for Nine Mile Point, Units 1 and 2, respectively, or until the NRC approves an applicable alternative in Regulatory Guide 1.147 or another document.
4.0 NRC STAFF EVALUATION The licensee referred to the EPRI report for the technical basis for the proposed alternative.
The major sections of the EPRI report that the licensee included in its submittal were Section 4, "Operating Experience"; Section 5, "Evaluation of Potential Degradation Mechanisms"; and Section 6, "Stress Analysis and Flaw Tolerance Evaluation."
The use of Sections 4 and 5 of the EPRI report regarding operating experience and potential degradation mechanisms has been found acceptable by the NRC staff, as documented in letter dated June 26, 2017 (ADAMS Accession No. ML17170A013), which approved similar relief requests for 19 units operated by Exelon (i.e., Exelon fleet-wide precedent). Therefore, the NRC staff focused its evaluation on the plant-specific applicability of the generic analyses contained in Section 6 of the EPRI report to Nine Mile Point, Units 1 and 2. The NRC staff notes that Nine Mile Point, Units 1 and 2 were 2 of the 19 units.
The licensee cited the NRC staff's approval of a similar request that used the generic stress analysis for the Vogtle Electric Generating Plant, Units 1 and 2, and the Joseph M. Farley Nuclear Plant, Unit 1 (i.e., Vogtle-Farley precedent). The NRC staff's review and approval is documented in a letter dated January 26, 2017 (ADAMS Accession No. ML17006A109).
Considering the approved precedent, the current evaluation focuses on the licensee's demonstration of plant-specific applicability of this generic stress analysis to the RPV flange threads for Nine Mile Point.
4.1 Stress Analysis The licensee referenced the generic stress analysis as part of its deterministic basis to support the proposed alternative. Stresses were determined from the finite element method analyses and used as input into the flaw tolerance analysis ( evaluated in Section 4.2 of this safety evaluation). In its submittal, the licensee summarized its plant-specific information in Table 1, "Comparison of Parameters to Values Used in Bounding Analysis," and Table 2, "RPV Flange Thread Geometry."
Table 1 of the submittal provides information on six key Nine Mile Point plant parameters:
number of studs, stud nominal diameter, RPV inside diameter at stud hole, flange thickness at stud hole, design pressure, and preload stress. This table shows that the stud nominal diameters for Nine Mile Point, Units 1 and 2, are higher than that used in the generic stress analysis, and the preload stresses for both units are less than the corresponding generic value, indicating that these two parameters are bounded by the generic analysis. Consequently, only the other four param~ters need to be evaluated. Three of the parameters are used to calculate the operating pressure load per stud through the following equation:
Load per stud= TT (design pressure) (RPV inside diameter at stud hole)2/
( 4 x number of studs)
The NRC staff verified the licensee's calculation and confirmed that the load per stud for Nine Mile Point, Units 1 and 2, is less than the corresponding generic value. Therefore, these three additional parameters are also bounded by the generic analysis.
The last parameter (flange thickness at stud hole) leaves less RPV flange material in front of the critical crack front for Nine Mile Point. Table 1 shows that the RPV flange thickness at the stud hole for Nine Mile Point, Units 1 and 2, is 13.84 and 13.5 inches, respectively, versus 16 inches for the generic stress analysis, and, therefore, is not bounded by the generic analysis. This feature is common to all boiling water reactors. Flange thickness at stud hole was evaluated in the NRC staff's review of a similar relief request for the James A. Fitzpatrick Nuclear Power Plant (FitzPatrick). The RPV flange thickness at FitzPatrick was 13.5 inches, which is the same as the lowest value for both units at Nine Mile Point. Therefore, the NRC staff's evaluation of FitzPatrick is applicable to both units at Nine Mile Point. The NRC staff concluded that the stress pattern of the generic analysis is not sensitive to"the reduced thickness between the stud hole and the outer edge of the RPV flange. The NRC staff's review and approval of the relief request for FitzPatrick (i.e., FitzPatrick precedent) is documented in letter dated May 30, 2018 (ADAMS Accession No. ML18039A854 ). Therefore, this last parameter for Nine Mile Point, Units 1 and 2, is also bounded by the generic analysis in the EPRI report.
Table 2 of the submittal provides information on the RPV flange thread geometry. For both units, the pitch is 8 threads per inch. For Nine Mile Point, Units 1 and 2, the thread depths are 0.06345 and 0.06700 inches, respectively. In the generic stress analysis, the corresponding values are 8 threads per inch and 0.06500 inches. As documented in the Exelon fleet-wide precedent, the NRC staff evaluated differences of this magnitude in thread geometry on the final stress intensity factor (K) results and concluded that the impact is negligible. The same conclusion applies to Nine Mile Point, Units 1 and 2.
As stated above, the preload stress and the load per stud due to pressure for Nine Mile Point, Units 1 and 2, are bounded by the generic analysis. In addition, the NRC staff found that the maximum heat-up rate for Nine Mile Point, Units 1 and 2, which is specified in Technical Specification (TS) 3.2.1, "Reactor Vessel Heatup and Cooldown Rates," is also bounded by the generic heat-up rate of 100 degrees Fahrenheit (°F) per hour. Therefore, all applied loads for Nine Mile Point, Units 1 and 2, are bounded by the generic loads. Based on this, the NRC staff determined that the generic stress analysis results in the EPRI report apply to Nine Mile Point, Units 1 and 2. However, since the driving force (i.e., the applied stress intensity factor (applied K)) of the flaw tolerance analysis depends on the component geometry and the postulated flaw shape, the effect of the reduced RPV flange thickness on the flaw tolerance evaluation needs to be addressed. This is evaluated in Section 4.2 of this safety evaluation.
4.2 Flaw Tolerance Evaluation The licensee referenced the flaw tolerance analysis in the EPRI report as part of its basis to support the proposed alternative. The flaw tolerance analysis in the EPRI report, including the crack growth analysis, is based on the principles of linear elastic fracture mechanics. Similar to evaluation of the stress analysis, the NRC staff's current evaluation of the flaw tolerance analysis focuses on the effect to the generic analysis results due to the Nine Mile Point, Units 1 and 2, RPV flange information: (1) the flange material property and the bolt-up temperature and (2) the reduced flange thickness.
Regarding the first effect, the licensee provides its RPV flange unirradiated nil ductility transition reference temperature (RT NoT) and bolt-up temperature values for both units in Table 4, "RPV Flange RT NDT and Bolt-Up Temperature," of its submittal. For Unit 1, the RPV flange RT NDT is 60 °F, and the bolt-up temperature is 70 °F to 120°F. For Unit 2, the RPV flange RT NDT is 10°F, and the bolt-up temperature is> 70 °F. The licensee stated that these values were determined using the RT NDT values from plant records. The NRC staff confirmed that this information is consistent with that in the safety evaluations approving relocation of the Nine Mile Point, Units 1 and 2, pressure-temperature limits from the technical specification to the licensee's controlled pressure-temperature limits report. Since preload is the dominant contributor to applied K, evaluation of the allowable K at the lowest pressure-temperature limits temperature is appropriate. Applying the minimum {T-RT NDT) of 1 O ° F for Unit 1 and 60 °F for Unit 2 to the fracture toughness (Kie) equation in ASME Code,Section XI, Appendix A, the NRC staff verified the licensee's calculated lower bound K,c value of 59 and 102 allowable stress intensity factor (ksivin) for Units and 1 and 2, respectively. Subsequently, applying the acceptance criteria of ASME Code,Section XI, IWB-3600 (with safety margin of v10), the NRC staff verified that the allowed applied K would be 18.5 and 32.3 ksi-vin for Units 1 and 2, respectively, which are both greater than all maximum K values in Table 3 for the preload case of the generic analysis.
Therefore, considering the first plant-specific information, the NRC staff determined that Nine Mile Point, Units 1 and 2, are bounded by the generic flaw tolerance analysis.
As documented in the FitzPatrick precedent, the NRC staff evaluated the effect of a smaller flange thickness on the applicability of the generic flaw tolerance evaluation. The NRC staff concluded that the generic flaw tolerance evaluation is still bounding, even when the thickness between the stud hole and the RPV flange outer edge is smaller than the generic model. Since the NRC staff's review at FitzPatrick evaluated the same flange thickness as the lowest value for Nine Mile Point, the NRC staff's conclusion for FitzPatrick is applicable to both units at Nine Mile Point.
Regarding use of the crack growth analysis in the EPRI report to support the proposed alternative, the NRC staff considers it acceptable, because the assumption of 400 occurrences of preload and 4,000 occurrences for heat-up/cool-down for an 80-year period in the generic analysis is conservative for Nine Mile Point, Units 1 and 2.
4.3 Licensee's Supplemental Bases The licensee described the maintenance activities it performs each time the RPV closure head is removed to detect and mitigate general degradation prior to returning the reactor to service.
Specifically, the licensee stated that the threads in the RPV flange are inspected for damage, cleaned, and lubricated prior to reinstallation of the RPV studs. The NRC staff considers these activities beneficial to flaw detection and could potentially reduce flaw initiation. Therefore, the conservative nature of the stress and flaw tolerance analyses is verified periodically and maintained.
5.0 CONCLUSION
Since the flaw tolerance analysis results meet the ASME Code,Section XI acceptance criterion, the NRC staff determined that the licensee's proposed alternative to the examination requirements for the stud-hole threads in RPV flange of the RPV for both Nine Mile Point, Units 1 and 2, provides an acceptable level of quality and safety. Thus, the NRC staff determines that the proposed alternative provides an acceptable level of quality and safety.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 1 O CFR 50.55a(z)( 1 ). Therefore, the NRC staff authorizes the request for alternative proposed in the licensee's submittal for the fifth 10-year ISi interval for Nine Mile Point, Unit 1, and the fourth 10-year ISi interval for Nine Mile Point, Unit 2.
All other requirements of Section XI of the ASME Code for which relief was not specifically requested and approved in the subject relief requests remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: Mark Yoo Dated: December 21, 2018
ML18334A236 OFFICE DORL/LPL 1 /PM DORL/LPL 1 /LA NAME MMarshall PTalukdar/LRonewicz DATE 12/06/2018 12/03/2018 DMLR/MVIB/BC*
DAIiey 11/27/2018