View source for L-MT-20-002, 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange

{{Adams
| number = ML20031E432
| issue date = 01/31/2020
| title = 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange
| author name = Church C
| author affiliation = Northern States Power Co, Xcel Energy
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000263
| license number = DPR-022
| contact person = 
| case reference number = L-MT-20-002
| document type = Letter type:L, Report, Technical
| page count = 20
}}

=Text=
{{#Wiki_filter:(l Xcel Energy*
RESP O N $ I 8 L E 8 Y H AT URE*                                 2807 West County Road 75 Monticello, MN 55362 January 31, 2020                                                                      L-MT-20-002 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 10 CFR 50.55a(z)(1) Request RR-015: Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange Pursuant to 10 CFR 50.55a(z)(1), Alternatives to codes and standards requirements, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests approval of an alternative to the examination requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange, on the basis that the proposed alternative provides an acceptable level of quality and safety.
Specifically, NSPM is requesting an alternative to the volumetric examination of the threads in the reactor pressure vessel flange that is required every interval. The basis for this alternative is provided in the enclosure.
The MNGP is currently operating in the fifth 10-year interval of the Inservice Inspection (ISI)
Program in compliance with the ASME Code, Section XI, 2007 Edition with the 2008 Addenda.
NSPM is submitting this 10 CFR 50.55a(z)(1) request for the remainder of the fifth 10-year ISI interval for MNGP, scheduled to end on May 31, 2022. NSPM requests approval of this alternative by February 28, 2021.
If there are any questions or if additional information is needed, please contact Mr. Richard Loeffler at (612) 342-8981 or Richard.Loeffler@xenuclear.com.

Document Control Desk Page2 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
                 *~
Christopher R. Church Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:    Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC

L-MT-20-002                                                                              NSPM Enclosure 10 CFR 50.55a(z)(1) Request RR-015 Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange

===1.0   ASME Code Components Affected===
Code Class:                        ASME Section XI Code Class 1 Examination Category:              B-G-1, Pressure Retaining Bolting, Greater than 2 in. (50 mm) in Diameter Item Number:                       B6.40, Threads in Flange

== Description:==
Reactor Vessel Threads in Flange 2.0   Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Rules for lnservice Inspection and Testing of Components of Light-Water Cooled Plants, 2007 Edition with the 2008 Addenda. The Monticello Nuclear Generating Plant (MNGP) Fifth 10-Year Inservice Inspection (ISI) Interval began on September 1, 2012, and is scheduled to end on May 31, 2022.
ASME Section XI, Appendix VIII requirements are implemented as required by, and as modified by, 10 CFR 50.55a. Procedures and personnel are qualified to the Performance Demonstration Initiative (PDI). The PDI Program document meets the requirements of 10 CFR 50.55a up through the 2013 Edition of ASME Section XI.
3.0   Applicable Code Requirement The Reactor [Pressure] Vessel (RPV) Threads in Flange, Examination Category B-G-1, Item Number B6.40, are examined using a volumetric examination technique with 100 percent of the flange threaded stud holes examined every ISI interval. The examination area is the one-inch area around each RPV stud hole, as shown on ASME Section XI Figure IWB-2500-12.
4.0   Proposed Alternative and Reason for the Request Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), Alternatives to codes and standards requirements, Northern States Power Company (NSPM) is requesting an alternative to the requirement under Section XI of the ASME Code, Examination Category B-G-1, Item Number B6.40, Threads in Flange, to perform in-service ultrasonic examinations of the RPV flange threads every interval. The Page 1 of 13

L-MT-20-002                                                                                 NSPM Enclosure proposed 10 CFR 50.55a(z)(1) alternative is to eliminate this inspection requirement.
The basis for elimination of these examinations is Electric Power Research Institute (EPRI) Technical Report (TR) No. 3002010354, entitled, Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements (Reference 1). Approval of this proposed alternative eliminates performance of unnecessary volumetric inspections of the RPV flange threads at the Monticello Nuclear Generating Plant (MNGP). This 10 CFR 50.55a(z)(1) alternative provides for more efficient conduct of plant refueling outages while maintaining an acceptable level of quality and safety.
Licensees within the United States and internationally have worked with EPRI and in 2017 produced the above referenced final report (TR No. 3002010354). The final report includes a survey of inspection results from 168 nuclear units that responded (including domestic and international units), a review of operating experience related to RPV flange / bolting, and a flaw tolerance evaluation. The conclusion from this evaluation was that these ASME Code Section XI examinations had not been identifying any service-induced degradation and the associated impact on worker exposure, personnel safety, critical path time, and additional time at reduced water inventory was not commensurate with performance.
Potential Degradation Mechanisms An evaluation of potential degradation mechanisms that could impact flange/threads reliability is described in the 2017 EPRI report. Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenomena, dealloying corrosion, general corrosion, stress relaxation, creep, mechanical wear, and mechanical / thermal fatigue. Other than the potential for mechanical / thermal fatigue, there are no active degradation mechanisms identified for the threads in flange component.
The final EPRI report also notes a general conclusion from ASME's Risk-Based Inspection: Development of Guidelines, (Reference 2) that when a component item has no active degradation mechanism present and a preservice inspection has confirmed that the inspection volume is in good condition (i.e., contains no flaws /
indications), then subsequent in-service inspections do not provide additional value going forward. As explained in the final EPRI report, the RPV flange ligaments have not only received the required pre-service examinations, but more than 10,000 ISIs have been carried out with no relevant findings.
To address the potential for mechanical / thermal fatigue, the EPRI report documents a stress analysis and flaw tolerance evaluation of the flange thread area to assess mechanical / thermal fatigue potential. The evaluation consists of two parts. In the first part, a stress analysis is performed considering all applicable loads on the threads in flange component. In the second part, the stresses at the critical locations of the component are used in a fracture mechanics evaluation to determine the allowable flaw Page 2 of 13

L-MT-20-002                                                                                NSPM Enclosure size for the component as well as how much time it will take for a postulated initial flaw to grow to the allowable flaw size using guidelines in the ASME Code, Section XI, IWB-3600, Analytical Evaluation of Flaws.
Stress Analysis As discussed in the EPRI report, a stress analysis was performed to determine the stresses at critical regions of the threads in flange component as input to a flaw tolerance evaluation. A bounding finite element model was developed to represent a typical threads in flange component for the fleet of nuclear plants. To create a representative geometry for the finite element model, a PWR design was selected as a representative geometry because of its higher design pressure and temperature. The largest RPV diameter of the PWRs was used along with the largest bolts and the highest number of bolts. The larger and more numerous bolt configuration results in less flange material between bolt holes, whereas the larger RPV diameter results in higher pressure and thermal stresses.
The details of the RPV parameters for MNGP as compared to the bounding values used in this evaluation are shown in Tables 1, 2, and 3. Table 1 provides a comparison of basic dimensions and loads. As this table shows, not all MNGP parameters are bounded by the parameters evaluated in the EPRI report; however, the preload stress for MNGP is bounded by that specified within the EPRI report. Specifically, the preload stress in the EPRI report is 42,338 psi whereas for MNGP the preload stress is 36,589 psi. Also, the design pressure (1,250 psi) for MNGP, which is a BWR, is considerably less than that of the PWR design pressure (2,500 psi) used in the analysis. Considering the preload stress and design pressure, the stress per stud for MNGP is smaller than that used in the analysis of the EPRI report. Therefore, the MNGP stress is bounded by the stress used in the analysis described in the EPRI report. The dimensions of the analyzed geometry are shown in Figure 1 in the attachment.
Table 2 provides a comparison of thread specifications. As this table shows, the flange hole diameter discussed within the EPRI report analysis is larger than the flange hole diameter at the MNGP. The larger hole diameter results in a smaller remaining ligament between stud holes and is therefore conservative. The pitch of the threads used in the analysis in the EPRI report is identical to the pitch of the threads at MNGP.
The depth of the threads at MNGP is slightly greater than that presented in the EPRI report analysis, resulting in lower stress in the threads at MNGP. Therefore, the MNGP configuration is bounded by the configuration used in the EPRI report analysis.
Table 3 provides a comparison of fracture toughness parameters between the controlling stud preload conditions used in the analysis and those at MNGP. As this table shows, since the Reference Temperature for Nil Ductility Transition (RTNDT) for the MNGP flange (10&deg;F), is much lower than the RTNDT used in the analysis (60&deg;F), the Page 3 of 13

L-MT-20-002                                                                                      NSPM Enclosure fracture toughness for the MNGP is much greater than that used in the analysis discussed in the EPRI report.
In summary, the comparisons shown in Tables 1, 2, and 3 demonstrate that the stress analysis discussed in the EPRI report is applicable to MNGP.
Table 1: Comparison of MNGP Parameters to Bounding Values Used in the Analysis Nominal                    Flange No. of    Minimum                 RPV Inside Diameter                 Thickness     Design    Preload Studs        No. of               Diameter at Plant                              of Flange                 at Flange   Pressure   Stress Currently      Studs                Flange Hole Hole                      Hole      (psig)      (psi)
Installed   Evaluated                 (inches)
(inches)                  (inches)
MNGP          64           64           6.0       206.375       10.625     1, 250    36,589 Bounding Values 54          NA           7.0         173           16        2,500    42,338 Used in Analysis Table 2: Comparison of MNGP Flange Thread Parameters to Bounding Values Used in the Analysis Flange Thread Flange Thread                             Flange Thread Depth Plant                                         Pitch Specification                                  (inches)
(thread/inch)
MNGP                  6-8UN-2B                  8                  0.067500 Bounding Values 7-8N-2B                  8                  0.065000 Used in Analysis Table 3: Comparison of MNGP Fracture Toughness Parameters to Bounding Values Used in the Analysis Flange Temp Flange RTNDT                         Flange T- RTNDT     Flange Fracture Plant                              During Bolt (oF)                                 (oF)         Toughness (ksiin)
Preload (oF)
MNGP                  10                 60                50                  89.6 Bounding Values Used              60                 60                 0                  53.9 in Analysis The analytical model for the stress analysis discussed in the EPRI report is shown in Figures 2 and 3 of the attachment. The loads considered in the analysis consisted of:
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L-MT-20-002                                                                                 NSPM Enclosure
* A design pressure of 2500 psig at an operating temperature of 600&deg;F was applied to all internal surfaces exposed to internal pressure.
* Bolt/stud preload - The preload on the bounding geometry is calculated as:
Ppreload = C x P x ID2 = 1.1 x 2500 x 1732 = 42,338 psi (291.9 MPa)
S x D2            54 x 62 Where, Ppreload =  Preload pressure to be applied on modeled bolt (psi)
P        =  Internal pressure (psi)
ID       =  Largest inside diameter of RPV (inch)
C        =  Bolt-up contingencies (+10%)
S        =  Least number of studs D        =  Smallest stud diameter (inch)
* Thermal stresses - The only significant transient affecting the bolting flange is heat-up/cooldown. This transient typically consists of a steady 100&deg;F/hour ramp up to the operating temperature with a corresponding pressure ramp up to the operating pressure.
The ANSYS finite element analysis program was used to determine the stresses in the threads in flange component for the three loads described above.
Flaw Tolerance Evaluation A flaw tolerance evaluation was performed using the results of the stress analysis described in the EPRI report to determine how long it would take an initial postulated flaw to reach the ASME Code, Section XI allowable flaw size. A linear elastic fracture mechanics evaluation consistent with ASME Code, Section XI, IWB-3600 was performed.
Stress intensity factors (Ks) at four flaw depths of 360&deg; inside-surface-connected, partial-through-wall circumferential flaws are calculated using finite element analysis (FEA) techniques with the model described above. The maximum stress intensity factor (K) values around the bolt hole circumference for each flaw depth (a) are extracted and used to perform the crack growth calculations. The circumferential flaw is modeled to start between the 10th and 11th flange threads from the top end of the flange because that is where the largest tensile axial stress occurs. The modeled flaw depth-to-wall thickness ratios (a/t) are 0.02, 0.29, 0.55, and 0.77, as measured in any direction from the stud hole. This creates an ellipsoidal flaw shape around the circumference of the flange as shown in Figure 4 of the attachment for the flaw model with a/t = 0.77 a/t crack Page 5 of 13

L-MT-20-002                                                                                   NSPM Enclosure model. The crack tip mesh for the other flaw depths follows the same pattern. When preload is not being applied, the stud, stud threads, and flange threads are not modeled. The model is otherwise unchanged between load cases.
The maximum K results are summarized in Table 4 for four crack depths. Because the crack tip varies in depth around the circumference, the maximum K from all locations at each crack size is conservatively used for the K-vs.-a profile.
Table 4: Maximum K vs. a/t K at Crack Depth (ksiin)
Load 0.02 a/t  0.29 a/t  0.55 a/t  0.77 a/t Case 1   Preload                            11.2      17.4      15.5      13.9 Case 2   Preload + Heatup + Pressure        13.0      19.8      16.1      16.3 As shown in Table 4, two load cases were considered in the evaluation.
: 1. Preload
: 2. Preload + Heatup + Pressure Case 1 involving only Preload is limiting since the operating temperature for this case is 60&deg;F while the operating temperature for Case 2 is approximately 550&deg;F. The value of (T - RTNDT) for Case 1 for the MNGP RPV flange is 50&deg;F as shown in Table 3.
From the equations in paragraph A-4200, Fracture Toughness, of the ASME Code, Section XI, Appendix A, the corresponding value of fracture toughness, KIc, is 89.6 ksiin for the MNGP flange. As discussed in the Safety Evaluation attached to Reference 4, for evaluation of a postulated flaw such as that considered in the analysis, a structural factor of 2, consistent with Section XI, Appendix G, Fracture Toughness Criteria for Protection Against Failure, of the ASME Code, can be used instead of the 10 for detected flaws in IWB-3610 of ASME Code, Section XI. If the structural factor of 2 is used, the allowable KIc is 44.8 ksiin. This is much greater than the maximum stress intensity factor (KI) for the bolt-up condition of 17.4 ksiin obtained in the analysis shown in Table 4, indicating considerable margin.
Page 6 of 13

L-MT-20-002                                                                                NSPM Enclosure As Table 4 shows, the allowable stress intensity factor is not exceeded for all crack depths up to the deepest analyzed flaw of a/t = 0.77. Hence the allowable flaw depth of the 360&deg; circumferential flaw is at least 77% of the thickness of the flange. The allowable flaw depth is assumed to be equal to the deepest modeled crack for the purposes of this analysis.
For the crack growth evaluation, an initial postulated flaw size of 0.2 inch (5.08 mm) is chosen consistent with the ASME Code, Section XI IWB-3500, Acceptance Standards, for flaws. The deepest flaw analyzed is a/t = 0.77 because of the inherent limits of the model. Two load cases are considered for fatigue crack growth: heatup/cooldown and bolt preload. The heat-up/cooldown load case includes the stresses due to thermal and internal pressure loads and is conservatively assumed to occur 50 times per year. The bolt preload is assumed to be present and constant during the load cycling of the heat-up/cooldown load case. The bolt preload load case is conservatively assumed to occur five times per year, and these cycles do not include thermal or internal pressure. The resulting crack growth was determined to be negligible due to the small delta K and the relatively low number of cycles associated with the transients evaluated. Because the crack growth is insignificant, the allowable flaw size will not be reached and the integrity of the component is not challenged for at least 80 years (original 40-year design life plus additional 40 years of plant life extension).
The stress analysis and flaw tolerance evaluation presented above show that the threads in flange component at MNGP is very flaw tolerant and can operate for 80 years without exceeding ASME Code, Section XI safety margins. This clearly demonstrates that the threads in flange component examinations can be eliminated without affecting the safety of the RPV.
Operating Experience Review Summary As discussed above, the results of the survey discussed in the EPRI report confirmed that the RPV threads in flange examinations are adversely impacting outage activities, such as worker exposure, personnel safety, and critical path time while not identifying any service induced degradation. Specifically, for the U.S. fleet, a total of 94 nuclear units have responded and none of these units have identified any type of degradation.
As can be seen in Table 5 below, the data is encompassing. The 94 units represent data from 33 BWRs and 61 PWRs. For the BWR units, a total of 3,793 examinations were conducted; and for the PWR units, a total of 6,869 examinations were conducted with no service induced degradation identified. The BWR response data includes information from all of the plant designs in operation in the United States and includes the BWR-2, -3, -4, -5, and -6 designs. The PWR response data includes the 2-loop, 3-loop, and 4-loop designs and each of the PWR NSSS (Nuclear Steam Supply System) designs (i.e., Babcock and Wilcox, Combustion Engineering, and Westinghouse).
Page 7 of 13

L-MT-20-002                                                                                NSPM Enclosure Table 5: Summary of Survey Results - United States Fleet Number of Number of Plant Type     Number of Units                           Reportable Examinations Indications BWR              33                 3,793                 0 PWR              61                 6,869                 0 Total            94                10,662                 0 Related RPV Assessments In addition to the examination history and flaw tolerance discussed above, the EPRI report discusses studies conducted in response to the issuance of the Anticipated Transient Without Scram (ATWS) Rule by the NRC. This rule was issued to require design changes to reduce expected ATWS frequency and consequences. Many studies have been conducted to understand the ATWS phenomena and key contributors to successful response to an ATWS event. In particular, the reactor coolant system (RCS) and its individual components were reviewed to determine weak links.
As an example, even though significant structural margin was identified in NRC SECY-83-293, Amendments to 10CFR50 Related to Anticipated Transients Without Scram (ATWS) Events, dated July 19, 1983, for PWRs, the ASME Service Level C pressure of 3200 psig was assumed to be an unacceptable plant condition. While a higher ASME service level might be defensible for major RCS components, other portions of the RCS could deform to the point of inoperability. Additionally, there was the concern that steam generator tubes might fail before other RCS components with a resultant bypass of containment. The key take-away for these studies is that the RPV flange ligament was not identified as a weak link and other RCS components were significantly more limiting. Thus, there is substantial structural margin associated with the RPV flange.
As the EPRI report discusses, the RPV threads in flange are performing with very high reliability, based on operating and examination experience. This is due to the robust design and a relatively benign operating environment (e.g., the number and magnitude of transients are small, and the threads are generally not in contact with primary water at plant operating temperatures/pressures, etc.). The robust design is manifested in that plant operation has been allowed at several plants, even with a bolt/stud assumed to be out of service. As such, significant degradation of multiple bolts/threads would be needed prior to any RCS leakage.
Page 8 of 13

L-MT-20-002                                                                                NSPM Enclosure Control of Non-Service Induced Degradation To protect against non-service related degradation, NSPM uses detailed procedures for RPV disassembly and reassembly to ensure protection and care of the studs and flange, including threads in flange. Tensioning and detensioning are performed in multiple passes. Protective covers and guide caps are used on the studs and the RPV flange. Removed studs and nuts are stored out of the vessel work area until needed for reassembly activities. Clearances and alignment are verified and observed to ensure stud damage does not occur when moving the RPV head and internals.
Threads, stud holes, and the flange surfaces are cleaned and threads are lubricated prior to reassembly. All activities are performed and documented in each step of their applicable procedures. These controlled maintenance activities provide further assurance that degradation is mitigated and detected prior to returning the reactor to service.
5.0   Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), Alternatives to codes and standards requirements, NSPM is requesting an alternative to the requirement under Section XI of the ASME Code, Examination Category B-G-1, Item Number B6.40, Threads in Flange, to perform in-service ultrasonic examinations of the RPV flange threads every interval. The proposed alternative is to eliminate this inspection requirement.
EPRI report TR No. 3002010354, Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements, provides the technical basis for the elimination of the RPV threads in flange examination requirement at the MNGP as described therein when coupled with the plant-specific information provided as the basis of this request. This report was developed because evidence had suggested that there have been no occurrences of service-induced degradation and there are negative impacts on worker exposure, personnel safety, and outage critical path time from performance of these examinations. Approval of this proposed alternative will eliminate performance of unnecessary volumetric inspections of the RPV flange threads at the MNGP.
Since there is reasonable assurance that the proposed alternative is an acceptable alternate approach to the performance of the ultrasonic examinations, NSPM requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that application of the alternative provides an acceptable level of quality and safety.
Page 9 of 13

L-MT-20-002                                                                                 NSPM Enclosure 6.0    Duration of the Proposed Alternative This 10 CFR 50.55a(z)(1) alternative will be applied for the duration of the inservice inspection interval defined in Section 2 of this request or until such time as the NRC approves an applicable alternative in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, or another document.
7.0    Precedent The NRC has authorized use of an alternate to examination of the RPV threads in flange for several utilities whose plants include both BWRs and PWRs based on the earlier version of the EPRI report (Reference 9). A partial list is provided below. Donald C. Cook Nuclear Plant, Units 1 and 2, (Reference 8), has been more recently authorized based upon the current version of the EPRI report. It should be noted that the MNGP is a BWR-3 design similar to several BWR-3s listed below (i.e., Dresden Units 2 and 3, and Quad Cities Units 1 and 2) for which alternate examinations have been authorized by the NRC.
Southern Nuclear Operating Company, Inc. (Reference 3)
* Vogtle Electric Generating Plant, Units 1and 2 (PWRs)
* Joseph M. Farley Nuclear Plant, Unit 1 (PWR)
Exelon Generation (Reference 4)
* Braidwood Station, Units 1 and 2 (PWR)
* Dresden, Units 2 and 3 (BWR)
* Byron Station Units, 1 and 2 (PWR)
* Quad Cities, Units 1 and 2 (BWR)
* Ginna (PWR)
* Peach Bottom, Units 2 and 3 (BWR)
* Three Mile Island, Unit 1 (PWR)
* Limerick, Units 1 and 2 (BWR)
* Calvert Cliff Nuclear Power Plant, Units
* Nine Mile Point, Units 1 and 2 (BWR) 1 and 2 (PWR)
* Clinton Power Station (BWR)
Duke Energy (Reference 5)
* Catawba Nuclear Station, Unit 2 (PWR)
* Brunswick Steam Electric Plant, Unit 1 (BWR)
* Shearon Harris Nuclear Power Plant,
* H. B. Robinson Steam Electric Plant, Unit 1 (PWR)                                       Unit 2 (PWR)
* McGuire Nuclear Station, Units 1 and 2
* Oconee Nuclear Station, Units 1, 2 (PWR)                                              and 3 (PWR)
Page 10 of 13

L-MT-20-002                                                    NSPM Enclosure Dominion Nuclear Connecticut, Inc. (References 6 and 7)
* Millstone Power Station, Units 2 and 3 (PWR)
* North Anna Power Station, Units 1 and 2 (PWR)
Indiana Michigan Power Company (Reference 8)
* Donald C. Cook Nuclear Plant, Units 1 and 2 (PWRs)
Page 11 of 13

L-MT-20-002                                                                              NSPM Enclosure 8.0   References
: 1. Electric Power Research Institute (EPRI) Technical Report (TR)
No. 3002010354, Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements, Final Report, dated December 2017
: 2. American Society of Mechanical Engineers, Risk-Based Inspection:
Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998
: 3. Letter from M. T. Markley (NRC) to C. R. Pierce (Southern Nuclear Operating Co. Inc.), Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 - Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads Inflange Inspection (CAC Nos. MF8061, MF8062, MF8070), dated January 26, 2017 (ADAMS Accession No. ML17006A109)
: 4. Letter from D. J. Wrona (NRC) to B. C. Hanson (Exelon Generating Company, LLC), Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant, Units 1 and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; Limerick Generation Station, Units 1 and 2; Nine Mile Point Nuclear Station, Units 1 and 2; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; R. E. Ginna Nuclear Power Plant; and Three Mile Island Nuclear Station, Unit 1 - Proposed Alternative to Eliminate Examination of Threads in Reactor Pressure Vessel Flange (CAC Nos. MF8712-MF8729 and MF9548),
dated June 26, 2017 (ADAMS Accession No. ML17170A013)
: 5. Letter from U. Shoop (NRC) to S. Capps (Duke Energy), Brunswick Steam Electric Plant, Unit No. 1; Catawba Nuclear Station, Unit No. 2; Shearon Harris Nuclear Power Plant, Unit No. 1; McGuire Nuclear Station, Unit Nos. 1 and 2; Oconee Nuclear Station, Unit Nos. 1, 2 and 3; and H. B. Robinson Steam Electric Plant, Unit No. 2 - Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads in Flange Inspection (CAC Nos. MF9513 - MF9521; EPID L-2017-LLR-0019, dated December 26, 2017 (ADAMS Accession No. ML17331A086)
: 6. Letter from J. G. Danna (NRC) to D. G. Stoddard (Dominion Nuclear Connecticut, Inc.), Millstone Power Station, Unit Nos. 2 and 3 - Alternative Requests RR-04-24 and IR-3-30 for Elimination of the Reactor Pressure Vessel Page 12 of 13

L-MT-20-002                                                                           NSPM Enclosure Threads in Flange Examination (CAC Nos. MF8468 and MF8469), dated May 25, 2017 (ADAMS Accession No. ML17132A187)
: 7. Letter from M. T. Markley (NRC) to D. G. Stoddard (Dominion), North Anna Power Station, Units 1 and 2 - Proposed Inservice Inspection Alternatives N1-14-NDE-009 and N2-14-NDE-004 (CAC Nos. MF9298 and MF9299; EPID L-2016-LLR-0018), dated December 6, 2017 (ADAMS Accession No. ML17132A663)
: 8. Letter from D. J. Wrona (NRC) to J. P. Gebbie (Indiana Michigan Power Company), Donald C. Cook Nuclear Plant, Units 1 and 2 - Proposed Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination EPID L-2018-LLR-0084), dated December 11, 2018, (ADAMS Accession No. ML18337A394)
: 9. EPRI TR No. 300200626, Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements, dated March 2016 (ADAMS Accession No. ML16221A068)
Page 13 of 13

ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a(z)(1) REQUEST RR-015 PROPOSED ALTERNATIVE FOR EXAMINATION OF ASME SECTION XI, EXAMINATION CATEGORY B-G-1, ITEM NUMBER B6.40, THREADS IN FLANGE FIGURES 1-4 (4 Pages Follow)

Figure 1 Modeled Dimensions R86.5J/                        ! 8 .5" 12.0"           I*
17 .Ou 7.0" 16.0" L
R83 .75" R4.5u
: 10. 75" R85 .69u I in. = 25.4 lllill

Figure 2 Finite Element Model Showing Bolt and Flange Connection ELENE11ITS RE.DiL NUM J>.ID Vessel Flanqe

Figure 3 Finite Element Model Mesh with Detail at Thread Location I
I I
.....-I 1
I I
1, I
I I
I I
I I
IIJ,

Figure 4 Cross Section of Circumferential Flaw with Crack Tip Element Inserted After 10th Thread from Top of Flange}}