RA-17-0016, Duke Energy - Proposed Alternative in Accordance with 10 CFR 50.55(z)(1) for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange (17-GO-001)
| ML17088A846 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Mcguire, Catawba, Harris, Brunswick, Robinson, McGuire  |
| Issue date: | 03/29/2017 |
| From: | Donahue J Duke Energy Carolinas, Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 17-GO-001, RA-17-0016 | |
| Download: ML17088A846 (13) | |
Text
(~ DUKE ENERGY~
Serial: RA-17-0016 March 29, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 DOCKET NO. 50-325 /RENEWED LICENSE NO. DPR-71 CATAWBA NUCLEAR STATION, UNIT NO. 2 DOCKET NO. 50-414 /RENEWED LICENSE NO. NPF-52 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400 I RENEWED LICENSE NO. NPF-63 MCGUIRE NUCLEAR STATION, UNIT NOS. 1AND2 JOSEPH DONAHUE Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-1758 Joseph.Donahue@duke-energy.com 10 CFR 50.55a DOCKET NOS. 50-369, 50-370 I RENEWED LICENSE NOS. NPF-9 AND NPF-17 OCONEE NUCLEAR STATION, UNIT NOS. 1, 2 AND 3 DOCKET NOS. 50-269, 50-270, 50-287 /RENEWED LICENSE NOS. DPR-38, DPR-47 AND DPR-55 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 I RENEWED LICENSE NO. DPR-23
SUBJECT:
PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(z)(1) FOR EXAMINATION OF ASME SECTION XI, EXAMINATION CATEGORY B-G-1, ITEM NUMBER B6.40, THREADS IN FLANGE (17-G0-001)
Ladies and Gentlemen:
Pursuant to 10 CFR 50.55a(z)(1), Duke Energy Carolinas, LLC and Duke Energy Progress, LLC (Duke Energy) hereby requests U.S. Nuclear Regulatory Commission's approval for an alternative to examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange. Enclosure 1 contains details regarding this request.
This submittal contains no new regulatory commitments. Duke Energy requests approval of this relief request by March 1, 2018, to support the Brunswick, Unit 1 refueling outage in March 2018 and the Harris, Unit 1 refueling outage in April 2018. If you have questions concerning this request, please contact Art Zaremba, Manager, Fleet Licensing, at (980) 373-2062.
Joseph Donahue Vice President - Nuclear Engineering
U.S. Nuclear Regulatory Commission RA-17-0016 Page 2
Enclosure:
- 1. Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange (17-GO-001) cc:
(all with Enclosure unless otherwise noted)
C. Haney, Regional Administrator USNRC Region II M. P. Catts, USNRC Resident Inspector - BSEP G. A. Hutto, III, USNRC Resident Inspector - MNS R. Patterson, USNRC Resident Inspector - HNP J. D. Austin, USNRC Resident Inspector - CNS E. L. Crowe, USNRC Resident Inspector - ONS J. Zeiler, USNRC Resident Inspector - RNP A. L. Hon, NRR Project Manager - BSEP M. C. Barillas, NRR Project Manager - HNP D. Galvin, NRR Project Manager - RNP S. Koenick, NRR Project Manager - ONS M. Mahoney, NRR Project Manager - CNS & MNS Mr. Cliff Dautrich, Bureau Chief North Carolina Department of Labor Boiler Safety Bureau 1101 Mail Service Center Raleigh, NC 27699-1101
Duke Energy Carolinas, LLC Duke Energy Progress, LLC Relief Request Serial #17-GO-001 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange
Relief Request Serial #17-GO-001 Page 2 of 11
- 1. ASME Code Component(s) Affected:
Reactor Vessels at plants identified in Table 1 of this request.
2. Applicable Code Edition and Addenda
The applicable Edition and Addenda of the ASME Code,Section XI is identified in Table 1.
Table 1 Plant/Unit(s)
ISI Interval ASME Section XI Code Edition/
Addenda Interval Start Date Interval End Date Brunswick Steam Electric Plant Unit 1 Fourth 2001 Edition, Through 2003 Addendum 5/11/2008 5/10/2018 Catawba Nuclear Station Unit 2 Fourth 2007 Edition, Through 2008 Addendum 8/19/2015 2/24/2026 Shearon Harris Nuclear Power Plant Unit 1 Third 2001 Edition, Through 2003 Addendum 5/2/2007 5/1/2018 Shearon Harris Nuclear Power Plant Unit 1 Fourth 2007 Edition, Through 2008 Addendum 5/2/20171 8/1/20271 McGuire Nuclear Station Units 1 and 2 Fourth 2007 Edition, Through 2008 Addendum 12/1/11 (Unit 1) 7/15/14 (Unit 2) 11/30/21 (Unit 1) 12/14/24 (Unit 2)
Oconee Nuclear Station Units 1, 2 & 3 Fifth 2007 Edition, Through 2008 Addendum 7/15/2014 7/15/2024 H.B. Robinson Steam Electric Plant Unit 2 Fifth 2007 Edition, Through 2008 Addendum 7/21/2012 7/30/2021
3. 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% 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 ASME Section XI Figure IWB-2500-12.
4. Reason for Request
In accordance with 10 CFR 50.55a(z)(1 ), Duke Energy Carolinas, LLC and Duke Energy Progress, LLC (Duke Energy) is requesting a proposed alternative to the requirement to perform in-service volumetric examinations of Examination Category B-G-1, Item Number B6.40, Threads in Flange. Duke Energy 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 Requirement (Reference 1), which provides the basis for elimination of the requirement. The report includes a survey of inspection results from over 1 The fourth inservice inspection interval plan has not yet been issued, so these interval dates are tentative. The fourth interval end date is proposed to be extended in accordance with IWA-2430(c).
Relief Request Serial #17-GO-001 Page 3 of 11 168 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 Reactor Coolant System 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 EPRI Report No. 3002007626 (Reference 1). Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenomena, dealloying corrosion, general and galvanic corrosion, stress relaxation, creep, mechanical wear, mechanical/thermal fatigue, and microbiologically induced corrosion (MIC). The EPRI report concluded that the only plausible degradation mechanisms for the threads in flange are thermal and mechanical fatigue.
The EPRI report notes a general conclusion from Reference 6, [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 inservice 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 preservice examinations and over 10,000 inservice 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.
Evaluation The evaluation consisted of two parts. In the first part, a stress analysis was performed considering all applicable loads on the threads in flange. In the second part, the stresses at the critical locations of the component were 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 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 threads in flange 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 is representative of the sixteen units identified in Tables 2-1 and 2-2 of Reference 1.
The details of the RPV parameters for Duke Energy plants, as compared to the bounding
Relief Request Serial #17-GO-001 Page 4 of 11 values used in the evaluation, are shown in Table 2.
Dimensions of the analyzed geometry are shown in Figure 1.
The analytical model is shown in Figures 2 and 3.
The loads considered in the analysis consisted of the following:
- A design pressure of 2500 psi at an operating temperature of 600°F was applied to all internal surfaces exposed to internal pressure.
- Bolt/stud preload stress - The stud preload on the bounding geometry was calculated using the equation shown below and the parameters listed in Table 2.
This equation was also used to calculate the unit-specific stud preload stresses listed in Table 2.
= 2 2
= (1.1 2500 1732) 54 62
= 42,338 Where:
Ppreload = Preload pressure to be applied on modeled bolt (psi)
P
= RPV Internal Design Pressure (psi)
ID
= Largest inside diameter of RPV (in)
C
= (in.) 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/cool down. 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.
Evaluation of Stress Analysis Results to the Duke Energy Plants in This Request The Duke Energy plants are bounded by the parameters evaluated in Reference 1 (as shown in Table 2), except as follows:
The Shearon Harris Nuclear Plant has a 14.69 inch flange thickness at the stud hole, compared to 16 inches used in the finite element analysis.
The Robinson Nuclear Plant has 50 studs, compared to the 54 studs used in the analysis to determine stud preload for the finite element analysis.
Duke Energy believes that the results of the stress analysis are valid for all of the Duke Energy plants for the following reasons:
- 1. The stress analysis bounds all of the unit-specific data for Catawba Unit 2, McGuire Units 1 and 2, and Oconee Units 1, 2, and 3.
- 2. The stress analysis results are conservative for Brunswick Unit 1, for reasons identified in the EPRI Report (lower design pressure and temperature than a PWR).
- 3. The computed stud preload stress for Shearon Harris is less than 74% of the 42,338 psi value used in the finite element analysis. The Shearon Harris Reactor
Relief Request Serial #17-GO-001 Page 5 of 11 Vessel stress report shows that the maximum range of primary plus secondary stress intensity in the flange is less than 50% of the allowable, and the flange cumulative usage factor for fatigue is only 0.0041. As a result, the EPRI analysis results are considered to be representative for the Shearon Harris Nuclear Plant.
- 4. The computed stud preload stress for Robinson is less than 63% of the 42,338 psi value used in the finite element analysis. The maximum range of primary plus secondary intensity in the Reactor Vessel flange is less than 75% of the allowable, and the flange cumulative usage factor for fatigue is only 0.007. As a result, the EPRI analysis results are considered to be representative for the Robinson Nuclear Plant.
Table 2 Comparison of Duke Energy Plant Parameters to Bounding Values Used in Analysis Plant No. of Studs Installed (S)
Stud Nominal Diameter (inches)
(D)
RPV Inside Diameter at Stud Hole (inches)
(ID)
Flange Thickness at Stud Hole (inches)
Design Pressure (psig)
(P)
Stud Preload Stress (psi)
Brunswick Steam Electric Plant Unit 1 (BWR) 64 6-3/42 220 11.59 1250 228222 Range of 6 BWR Units in EPRI Report 56-76 6
217-250 12.9-14.5 1250 N/A Catawba Nuclear Station Unit 2 (PWR) 54 7
172.50 16.25 2485 30740 Shearon Harris Nuclear Power Plant Unit 1 (PWR) 58 6
154.62 14.69 2485 31298 McGuire Nuclear Station Unit 1 (PWR) 54 7
172.56 16.22 2485 30762 McGuire Nuclear Station Unit 2 (PWR) 54 7
172.50 16.25 2485 30740 Oconee Nuclear Station Unit 1 (PWR) 60 6-1/2 167.50 16.06 2500 30436 Oconee Nuclear Station Unit 2 (PWR) 60 6-1/2 167.50 16.06 2500 30436 Oconee Nuclear Station Unit 3 (PWR) 60 6-1/2 167.50 16.06 2500 30436 H.B. Robinson Steam Electric Plant Unit 2 (PWR) 50 7
154.50 17.22 2485 26632 Parameters Used to Determine Stud Preload Stress in Finite Element Analysis in EPRI Report:
54 6
173.00 N/A 2500 Parameters Used in Finite Element Analysis Model in EPRI Report:
60 7
173.00 16.00 Flaw Tolerance Evaluation The EPRI Report (Reference 1) includes a flaw tolerance evaluation using the results of the stress analysis in that report. The evaluation 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 for 360° inside-surface-connected, part-through-wall circumferential flaws were calculated using finite element analysis techniques with the model described in Reference 1. The maximum stress intensity factor (K) values 2 Based on Bushing Diameter (O.D.)
Relief Request Serial #17-GO-001 Page 6 of 11 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 for the flaw model with a/t = 0.77 a/t 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 3 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/t profile.
Table 3 Maximum K vs. a/t Load K at Crack Depth (ksiin) 0.02 a/t 0.29 a/t 0.55 a/t 0.77 a/t Preload 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-3612 for normal operating conditions:
KI < KIc/10 = 69.6 ksiin
- Where, KI = Applied stress intensity factor (ksiin)
KIc = Lower bound fracture toughness at operating temperature (220 ksiin)
As shown in Table 3, 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° 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 in. (5.08 mm) was chosen consistent with the ASME Code,Section XI IWB-3500 flaw acceptance standards.
The deepest flaw analyzed is a/t = 0.77 because of the inherent limits of the model.
Two load cases were considered for fatigue crack growth: heat-up/cool down and bolt preload. The heat-up/cool down load case included the stresses due to thermal and internal pressure loads and was conservatively assumed to occur 50 times per year. The bolt preload was assumed to be present and constant during the load cycling of the heat-up/cool down load case. The bolt preload load case was 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 would not be reached and the integrity of the component would not be challenged for at least 80 years (original 40-year design life plus
Relief Request Serial #17-GO-001 Page 7 of 11 additional 40 years of plant life extension).
The stress analysisIflaw tolerance evaluation presented above demonstrates that the RPV threads in flange at the units in this request are very flaw tolerant, clearly demonstrating that the thread in flange examinations can be eliminated without affecting the safety of the RPV.
Operating Experience Review Summary The EPRI Report (Reference 1) concludes that the examinations of RPV threads in flange are adversely impacting outage activities (worker exposure, personnel safety, radwaste, critical path time, and additional time at reduced water inventory) and have not identified any service-induced degradation. For the U.S. fleet, a total of 94 units provided input to the EPRI Report, and none of these units have identified any type of degradation. As shown in Table 4 below (reproduced from Table 3-1 of Reference 1), not a single unit has reported detecting a reportable indication in more than 10,000 examinations conducted. The 94 units identified in Table 4 represent data from 33 BWRs and 61 PWRs. No service-induced degradation was identified in 3,793 BWR examinations and 6,869 PWR examinations. The response data included information from all of the plant designs in operation in the U.S.,
including BWR-2, -3, -4, -5 and -6 designs, as well as PWR 2-loop, 3-loop and 4-loop designs and each of the PWR NSSS designs (i.e., Babcock & Wilcox, Combustion Engineering and Westinghouse).
Table 4: Summary of Survey Results - US Fleet Plant Type Number of Units Number of Examinations Number of 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 ATWS frequency and consequences. Many studies have been conducted to understand the ATWS phenomena and key contributors to successful response to an ATWS event. Reference 1 indicates that 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 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
Relief Request Serial #17-GO-001 Page 8 of 11 components were significantly more limiting. Thus, there is substantial structural margin associated with the RPV flange.
In summary, Reference 1 concludes that the RPV threads in flange examination can be eliminated without increasing plant risk or posing any safety concerns for the RPV.
5. Proposed Alternative and Basis for Use
In lieu of the in-service volumetric examination requirement, Duke Energy proposes that this request, including the EPRI 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.
The EPRI 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 B6.40). 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 RCS water inventory.
For these reasons, Duke Energy requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that the alternative provides an acceptable level of quality and safety.
6. Duration of Proposed Alternative
This alternative is requested for the duration of the inservice inspection intervals listed in Table 1 of this request.
- 7. Precedent:
The NRC has recently authorized the use of an alternative to examination of the RPV threads in flange for the Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M.
Farley Nuclear Plant, Unit 1, based on Reference 1 and site specific analysis. The NRC approval was provided as noted in Reference 2.
- 8.
References:
- 1.
Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements. EPRI, Palo Alto, CA: 2016. 3002007626, Technical Update, March 2016 (ADAMS Accession No. ML16221A068)
- 2.
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 in Flange Inspection (CAC Nos. MF8061, MF8062, MF8070) dated January 26, 2017 (ADAMS Accession No. ML17006A109)
- 3.
Exelon Generation Co. LLC, Proposed Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange, dated October 31, 2016, (ADAMS Accession No. ML16306A270)
- 4.
Virginia Electric and Power Company (Dominion) North Anna Power Station, Units 1 and 2, ASME Section XI Inservice Inspection Program Request for Proposed Alternative N1-14-NDE-009 and N2-14-NDE-004, dated November 30, 2016 (ADAMS Accession No. ML16340B092)
Relief Request Serial #17-GO-001 Page 9 of 11 5.
Dominion Nuclear Connecticut, Inc. Millstone Power Station Units 2 and 3 - Proposed Alternative Requests RR-04-24 an 1R-3-30 for Elimination of the Reactor Pressure Vessel Threads in Flange Examination, dated October 6, 2016, (ADAMS Accession No. ML16287A724) 6.
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.
Relief Request Serial #17-G0-001 Figure 1 Modeled Dimensions R86.5" 12.0" 17.0" 7.0"
..L 16.0' I R83.75"..
_Q.1.5_" _
R85.69" Figure 2 8.5" R4.5" Reference 1 Figure 6-2 Finite Element Model Showing Bolt and Flange Connection Vessel_Flange Reference 1 Figure 6-3 Page 10of11
Relief Request Serial #17-GO-001 Page 11 of 11 Figure 3 Finite Element Model Mesh with Detail at Thread Location Figure 4 Cross Section of Circumferential Flaw with Crack Tip Element Inserted After 10th Thread from Top of Flange Reference 1 Figure 6-4 Reference 1 Figure 6-8