RS-16-142, Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange
| ML16306A270 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, Dresden, Peach Bottom, Nine Mile Point, Byron, Braidwood, Limerick, Ginna, Quad Cities, Crane  |
| Issue date: | 10/31/2016 |
| From: | Jim Barstow Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-16-142 | |
| Download: ML16306A270 (17) | |
Text
200 Exelon Way Exelon Generation Kennett Square. PA 19348 www.exeloncorp.com RS-16-142 October 31, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 10 CFR 50.55a Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 N RC Docket Nos. STN 50-454 and STN 50-455 Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-41 O Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244
U.S. Nuclear Regulatory Commission Proposed Alternative for Examination Threads in Flange October 31, 2016 Page2 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289
Subject:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange In accordance with 10 CFR 50.55a(z)(1 ), Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components, 11 on the basis that the proposed alternative provides an acceptable level of quality and safety. Specifically, Exelon is requesting an alternative to volumetric examination of reactor vessel threads in closure head flange connections.
The basis for this request is provided in the Attachment. There are no commitments contained in this submittal.
If you have any questions regarding this submittal, please contact Stephanie Hanson at 610-765-5143.
Respectfully,
~~~
James Barstow Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC
Attachment:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange cc:
Regional Administrator - NRC Region I Regional Administrator - NRC Region Ill NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station NRC Senior Resident Inspector - Calvert Cliffs Nuclear Power Plant NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - Limerick Generating Station NRC Senior Resident Inspector - Nine Mile Point Nuclear Station NRC Senior Resident Inspector - Peach Bottom Atomic Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station NRC Senior Resident Inspector - R.E. Ginna Nuclear Power Plant NRC Senior Resident Inspector - Three Mile Island Nuclear Station, Unit 1 NRC Project Manager - Braidwood Station NRC Project Manager - Byron Station NRC Project Manager - Calvert Cliffs Nuclear Power Plant NRC Project Manager - Dresden Nuclear Power Station
U.S. Nuclear Regulatory Commission Proposed Alternative for Examination Threads in Flange October 31, 2016 Page3 cc (contd.):
NRC Project Manager - Limerick Generating Station NRC Project Manager - Nine Mile Point Nuclear Station NRC Project Manager - Peach Bottom Atomic Power Station NRC Project Manager - Quad Cities Nuclear Power Station NRC Project Manager - R.E. Ginna Nuclear Power Plant NRC Project Manager - Three Mile Island Nuclear Station, Unit 1 S. Gray, MD, DNR
ATTACHMENT Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange
10 CFR 50.SSa RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 1 of 13)
- 1. ASME Code Component(s) Affected:
All American Society of Mechanical Engineers (ASME),Section XI, Examination Category 8-G-1, Item Number 86.40 threads in flange locations at the sites listed in Section 2 of this relief request.
2. Applicable Code Edition and Addenda
PLANT INTERVAL EDITION START END Braidwood Station, Third 2001 Edition, through July 29, 2008 July 28, 2018 Units 1 and 2 2003 Addenda October 17, 2008 October 16, 2018 Byron Station, Fourth 2007 Edition, through July 16, 2016 July 15, 2025 Units 1 and 2 2008 Addenda Calvert Cliffs Nuclear Power Fourth 2004 Edition October 10, 2009 June 30, 2019 Plant, Units 1 and 2 Dresden Nuclear Power Fifth 2007 Edition, through January 20, 2013 January 19, 2023 Station, Units 2 and 3 2008 Addenda Limerick Generating Station, Fourth 2007 Edition, through February 1, 2017 January 31, 2027 Units 1 and 2 2008 Addenda Nine Mile Point Nuclear Fourth 2004 Edition August23,2009 August22,2019 Station, Unit 1 Nine Mile Point Nuclear Third 2004 Edition April 5, 2008 June 15, 2018 Station, Unit 2 Peach Bottom Atomic Power Fourth 2001 Edition, through November 5, 2008 December 31, 2018 Station, Units 2 and 3 2003 Addenda Quad Cities Nuclear Power Fifth 2007 Edition, through April 2, 2013 April 1, 2023 Station, Units 1 and 2 2008 Addenda R.E. Ginna Nuclear Power Fifth 2004 Edition January 1, 2010 December 31, 2019 Plant Three Mile Island Nuclear Fourth 2004 Edition April 20, 2011 April 19, 2022 Station, Unit 1
3. Applicable Code Requirement
The Reactor Pressure Vessel (RPV) threads in flange, Examination Category 8-G-1, Item Number 86.40, are examined using a volumetric examination technique with 100%
of the flange threaded stud holes examined every In-service Inspection (ISi) interval.
The examination area is the one-inch area around each RPV stud hole, as shown on Figure IWB-2500-12.
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 2 of 13)
4. Reason for Request
In accordance with 10 CFR 50.55a(z)(1 ), Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative from the requirement to perform in-service ultrasonic examinations of Examination Category B-G-1, Item Number 86.40, Threads in Flange.
Exelon has worked with the industry to evaluate eliminating the RPV threads in flange examination requirement. Licensees in the U.S. and internationally have worked with the Electric Power Research Institute (EPRI) to produce Technical Report No. 3002007626, "Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements" (Reference 1 ), which provides the basis for elimination of the requirement.
The report includes a survey of inspection results from over 168 units, a review of operating experience related to RPV flange/bolting, and a flaw tolerance evaluation. The conclusion from this evaluation is that the current requirements are not commensurate with the associated burden (worker exposure, personnel safety, radwaste, critical path time, and additional time at reduced water inventory) of the examination. The technical basis for this alternative is discussed in more detail below.
Potential Degradation Mechanisms An evaluation of potential degradation mechanisms that could impact flange/threads reliability was performed as part of Reference 1. Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenome, dealloying corrosion and 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 EPRI report notes a general conclusion from Reference 2, (which includes work supported by the U.S. Nuclear Regulatory Commission (NRC)) 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., no flaws I indications), then subsequent in-service inspections do not provide additional value going forward. As discussed in the Operating Experience review summary below, the RPV flange ligaments have received the required pre-service examinations and over 10,000 in-service inspections, with no relevant findings.
To address the potential for mechanical/thermal fatigue, Reference 1 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 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-3500. The Pressurized Water Reactor (PWR) design was selected because of its higher design
10 CFR 50.SSa RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 3 of 13) pressure and temperature. A representative geometry for the finite element model used the largest PWR RPV diameter 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.
Stress Analysis A stress analysis was performed in Reference 1 to determine the stresses at critical regions of the thread in flange component as input to a flaw tolerance evaluation. Sixteen nuclear plant units (ten PWRs and six Boiling Water Reactors (BWRs)) were considered in the analysis. The evaluation was performed using a geometric configuration that bounds the sixteen units considered in this effort. The details of the RPV parameters for Exelon plants as compared to the bounding values used in the evaluation are shown in Table 1. Not all the Exelon plants are bounded by the parameters evaluated in Reference 1; however, the preload stresses for each unit are bounded by the Reference 1 report. Specifically, the Reference 1 preload stress is 42,338 psi whereas the maximum Exelon Unit preload stress is 36,711 psi at Limerick Units 1 and 2. The Exelon unit specific stresses are bounded by the Reference 1 report which demonstrates that the report remains applicable to all the units identified in this relief request. Dimensions of the analyzed geometry are shown in Figure 1.
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 4 of 13)
Table 1: Comparison of Exelon Plant Parameters to Bounding Values Used in Analysis No. of Minimum Stud RPV Inside Flange Studs No. of Nominal Diameter Thickness Design Pre load Plant Currently Studs Diameter at Stud at Stud Hole Pressure Stress Hole (psia)
(psi)
Installed Evaluated (inches)
(inches)
(inches)
Braidwood 1 54 53 6.75 171.06 16.97 2500 33,323 Braidwood 2 53 53 6.75 171.06 16.97 2500 33,323 Byron 1 54 53 6.75 171.06 16.97 2500 33,323 Byron 2 54 53 6.75 171.06 16.97 2500 33,323 Calvert Cliffs 54 54 7
172 16.5 2500 30,747 1
Calvert Cliffs 54 54 7
172 16.5 2500 30,747 2
Dresden 2 92 92 6
251.37 13.94 1265 26,549 Dresden 3 92 92 6
251.37 13.94 1265 26,549 Ginna 48 48 6
128.31 14.56 2500 26,202 Limerick 1 76 76 5.62 251.87 12.56 1265 36,711 Limerick 2 76 76 5.62 251.87 12.56 1265 36,711 Nine Mile 64 64 6.25 213.44 13.84 1265 25,356 Point 1 Nine Mile 76 76 6.5 251.5 13.5 1265 27,411 Point 2 Peach 92 92 6
267.25 14 1280 30,363 Bottom 2 Peach 92 92 6
267.25 14 1280 30,363 Bottom 3 Quad Cities 1 92 92 6
251.37 13.94 1265 26,542 Quad Cities 2 92 92 6
251.37 13.94 1265 26,542 Three Mile 60 60 6.5 167.25 16.06 2515 30,527 Island 1 Range for 16 Units 54-60 54 6.5 - 7.0 157 - 173 15 -16 2500 42,338 Considered Bounding Values Used 54 NA 6.0 173 16 2500 NA in Analysis The analytical model is shown in Figures 2 and 3. The loads considered in the analysis consisted of:
A design pressure of 2500 psia at an operating temperature of 600°F was applied to all internal surface exposed to internal pressure.
Bolt/stud preload - The preload on the bounding geometry is calculated as:
C
- P
- ID2 1.1*2500* 173 2
Ppreload = ----
S
- D2
--5-4-. - = 42,338 psi
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange
{Page 5 of 13) where:
Ppreload
=
Preload pressure to be applied on modeled bolt (psi) p
=
Internal pressure (psi)
ID
=
Largest inside diameter of RPV (in.)
c
=
Bolt-up contingencies (+10%)
s
=
Least number of studs D
=
Smallest stud diameter (in.)
Thermal stresses - The only significant transient affecting the bolting flange is heat-up/cooldown. This transient typically consists of a steady 100°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 thread 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 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 (K's) at four flaw depths of a 360° inside-surface-connected, partial-through-wall circumferential flaws are calculated using finite element analysis 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 1 Oth 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 (alt) 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 for the flaw model with alt= 0.77 alt crack 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 2 for the 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.
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 6 of 13)
Table 2: Maximum K vs. alt Load Kat Crack Depth (ksb/in) 0.02 alt 0.29 alt 0.55 alt 0.77 alt Pre load 11.2 17.4 15.5 13.9 Preload + Heatup + Pressure 13.0 19.8 16.1 16.3 The allowable stress intensity factor was determined based on the acceptance criteria in ASME Section XI, IWB-361 O/Appendix A which states that:
K1 < K1J...J 1 O = 69.6 ksi...Jin
- Where, K1 =Allowable stress intensity factor (ksi...Jin)
K1c = Lower bound fracture toughness at operating temperature {220 ksi...Jin)
As can be seen from Table 2, the allowable stress intensity factor is not exceeded for all crack depths up to the deepest analyzed flaw of alt= 0.77. Hence the allowable flaw depth of the 360° 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.
Some Exelon plants have RPV closure heads in service without studs in all the original designed flange bolt hole locations. The thread in flange configuration has much redundancy. As seen from the stress intensity factor (K) calculation documented in Table 6-1 of Reference 1 (reproduced in Table 2 above), the maximum K is 19.8 ksi...Jin. The allowable K calculated in Section 6.2.2 of the report is 69.6 ksi...Jin, significantly higher than the calculated value. Assuming an RPV flange with 60 studs originally and one inoperable stud, the increase in K is about 1. 7% resulting in a maximum K of about 20.14 ksi...Jin which is still significantly less than the allowable value.
For the crack growth evaluation, an initial postulated flaw size of 0.2 in. (5.08 mm) is chosen consistent with the ASME Code,Section XI IWB-3500 flaw acceptance standards. The deepest flaw analyzed is alt = 0. 77 because of the inherent limits of the model. Two load cases are considered for fatigue crack growth: heat-up/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 Kand 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).
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 7 of 13)
The stress analysis I flaw tolerance evaluation presented above shows that the thread in flange component at the units in the relief request is very flaw tolerant and can operate for 80 years without violating ASME Code,Section XI safety margins. This clearly demonstrates that the thread in flange examinations can be eliminated without affecting the safety of the RPV.
Operating Experience Review Summary As discussed above, the results of the survey, which includes results from the Exelon plants in this relief request, confirmed that the RPV threads in flange examination are adversely impacting outage activities (worker exposure, personnel safety, radwaste, critical path time, and additional time at reduced water inventory) while not identifying any service induced degradations. Specifically, for the U.S. fleet, a total of 94 units have responded to date and none of these units have identified any type of degradation. As can be seen in Table 3 below, the data is encompassing. The 94 units represent data from 33 BWRs and 61 PW Rs. For the BWR units, a total 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 response data includes information from all of the plant designs in operation in the U.S. and includes BWR-2, -3, -4, -5 and -6 designs. The PWR plants include the 2-loop, 3-loop and 4-loop designs and each of the PWR NSSS designs (i.e., Babcock &
Wilcox, Combustion Engineering and Westinghouse).
Table 3: Summary of Survey Results - US Fleet Number of Number of Number of Plant Type Units Examinations Reportable 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, Reference 1 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 A TWS frequency and consequences. Many studies have been conducted to understand the A TWS phenomena and key contributors to successful response to an A TWS 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 USNRC SECY-83-293 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
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 8 of 13) 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.
In summary, Reference 1 identifies that 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 is small, 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.
5. Proposed Alternative and Basis for Use
In lieu of the in-service requirements for a volumetric ultrasonic examination, Exelon proposes that the industry report (Reference 1) provides an acceptable technical basis for eliminating the requirement for this examination because the alternative maintains an acceptable level of quality and safety.
This report provides the basis for the elimination of the RPV threads in flange examination requirement (ASME Section XI Examination Category B-G-1, Item Number 86.40). 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 dose, personnel safety, radwaste, critical path time for these examinations and additional time at reduced water inventory.
Since there is reasonable assurance that the proposed alternative is an acceptable alternate approach to the performance of the ultrasonic examinations, Exelon requests authorization to use the proposed alternative pursuant to 1 O CFR 50.55a(z)(1) on the basis that use of the alternative provides an acceptable level of quality and safety.
To protect against non-service related degradation, Exelon uses detailed procedures for the care and visual inspection of the RPV studs and the threads in flange each time the RPV closure head is removed. Care is taken to inspect the RPV threads for damage and to protect threads from damage when the studs are removed. Prior to reinstallation, the studs and stud holes are cleaned and lubricated. The studs are then replaced and tensioned into the RPV flange. This activity is performed each time the closure head is removed, and the procedure documents each step. These controlled maintenance activities provide further assurance that degradation is detected and mitigated prior to returning the reactor to service.
6. Duration of Proposed Alternative
This relief request will be applied for the duration of the inservice inspection intervals defined in Section 2 of this relief request or such time as the NRC approves an applicable alternative in Regulatory Guide 1.147 or other document.
10 CFR 50.SSa RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange
{Page 9 of 13)
- 7. Precedent:
None
- 8.
References:
- 1. Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements. EPRI, Palo Alto, CA: 2016. 3002007626. (ADAMS Accession No. ML16221A068)
- 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.
10 CFR 50.SSa RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 10 of 13)
R86.5" 17.0
R83. 75" R85.69" Figure 1 Modeled Dimensions 12.0" 7.0" 16.0"
- 10. 75" 8.5" R4.5"
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 11 of 13)
Figure 2 Finite Element Model Showing Bolt and Flange Connection ANO_Vessel_Flange
10 CFR 50.SSa RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 12 of 13)
Figure 3 Finite Element Model Mesh with Detail at Thread Location
10 CFR 50.55a RELIEF REQUEST:
Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange (Page 13 of 13)
Figure 4 Cross Section of Circumferential Flaw with Crack Tip Elements Inserted After 10th Thread from Top of Flange