AEP-NRC-2018-33, Proposed Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination

Proposed Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination

ML18169A148
Person / Time
Site:	Cook
Issue date:	06/14/2018
From:	Lies Q Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2018-33
Download: ML18169A148 (14)

Text

m INDIANA MICHIGAN Indiana Michigan Power POWER* One Cook Place Bridgman, Ml 49106 A unit ofAmerican Electric Power lndianaMichiganPower.com June 14, 2018 AEP-NRC-2018-33 10 CFR 50.55a Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington , DC 20555-0001 Donald C. Cook Nuclear Plant, Units 1 and 2 Proposed Alternative Request for Elim ination of the Reactor Pressure Vessel Threads in Flange Examination Pursuant to 10 CFR 50.55a(z)(1 ), Indiana Michigan Power Company (l&M) , the licensee for Donald C. Cook Nuclear Plant (CNP) , hereby requests U. S. Nuclear Regulatory Commission (NRC) approval for an alternative to the examination requirements of ASME Section XI , Examination Category B-G-1 , Item Number B6.40, Threads in Flange. The proposed alternative is provided in the enclosure to this letter.

l&M would like to request NRC review and approval of the proposed alternative by December 31 , 2018, to support the next CNP Unit 1 refueling outage currently scheduled to occur during spring 2019.

There are no new or revised regulatory commitments made in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely,

)~J.t

a. C e Lies Site Vice President Indiana Michigan Power Company BMC/mll

U. S. Nuclear Regulatory Commission AEP-NRC-2018-33 Page 2

Enclosure:

10 CFR 50 .55a Relief Request Number ISIR-4-07, Proposed Alternative to ASME Section XI for Elimination of Reactor Vessel - Threads in Flange Examination c: R. J. Ancon a, MPSC A W . Dietrich, NRC , Washington , DC MDEQ - RMD/RPS NRC Resident Inspector K. S. West, NRC, Region Ill A J. William son , AEP Ft. Wayne , w/o enclosures

Enclosure to AEP-NRC-2018-33 10 CFR 50.55a Relief Request Number ISIR-4-07 Proposed Alternative to ASME Section XI for Elimination of Reactor Vessel - Threads in Flange Examination

1. ASME Code Component(s) Affected The affected components are the reactor vessels at Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2.

2. Applicable Code Edition and Addenda

The applicable code edition is the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI (ASME Section XI) , 2004 edition , no addenda (Reference 1).

3. Applicable Code Requirement

The Reactor Pressure Vessel (RPV) threads in flange are required to be examined using a volumetric examination technique with 100% of the flange ligament areas examined every inservice inspection (ISi) interval. The examination area is the one inch area around each RPV stud hole, as shown in ASME Section XI (Reference 1), Figure IWB-2500-12.

4. Reason for Request

In accordance with 10 CFR 50.55a(z)(1) , Indiana Michigan Power Company (l&M) 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. Licensees in the United States (U . S.) and internationally have worked with the Electric Power Research Institute (EPRI) to produce Technical Report No. 3002010354, "Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements" (referred to herein as the EPRI report or Reference 2) ,

which provides the basis for elimination of the requirement. The report includes a survey of inspection results from 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 cou ld impact flange/thread reliability was performed as part of the EPRI report (Reference 2) . Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue , stress corrosion cracking, microbiologically induced corrosion , velocity phenomena, de-alloying corrosion, general and

Enclosure to AEP-NRC-2018-33 Page 2 galvanic corrosion , stress relaxation, creep, mechanical wear, and mechanical and thermal fatigue . The EPRI report (Reference 2) concluded that the only plausible degradation mechanisms for the threads in flange are thermal and mechanical fatigue.

The EPRI report (Reference 2) notes a general conclusion from Risk-Based Inspection :

Development of Guidelines, Volume 2 (Reference 3) , which includes work supported by the U. S. Nuclear Regulatory Comm ission (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/indications) , then subsequent ISi do not provide additional value going forward . As discussed in the Operating Experience review summary below, the reported industry RPV flange ligaments have received the required preservice examinations and over 10,000 ISi , with no relevant findings.

To address the potential for mechanical/thermal fatigue, Reference 2 documents a stress analysis and flaw tolerance evaluation of the flange thread area to assess mechanical/thermal fatigue potential.

Evaluation The evaluation (Reference 2) 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 Subsection IWB-3500 of ASME Section XI (Reference 1). 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 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 The stress analysis performed and documented in Reference 2 determined the stresses at critical regions of the threads in flange as input to a flaw tolerance evaluation . Sixteen nuclear plants (ten PWRs and six Boiling Water Reactors (BWRs)) were considered in the stress analysis in Reference 2. 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 2.

The details of the RPV parameters for CNP Units 1 and 2, as compared to the bounding values used in the evaluation , are shown in Table 1. As shown in Table 1 below, the diameter of the stud used in the analysis is smaller than that at CNP Units 1 and 2, and the inside diameter of the RPV in the analysis is larger than that at CNP Units 1 and 2. The number of studs installed is the same. The smaller stud diameter results in higher preload stresses per bolt, and the

Enclosure to AEP-NRC-2018-33 Page 3 larger RPV inside diameter results in higher pressure and thermal stresses. Therefore, the stresses from the analyzed configuration would be conservative in application to CNP Units 1 and 2. The dimensions of the analyzed geometry are shown in Figure 1 at the end of this Enclosure.

Table 1: Comparison of CNP Plant Parameters to Bounding Values Used in Analysis Stud RPV Inside Flange No. of Design Nominal Diameter at Thickness Plant Studs Pressure Diameter Stud Hole at Stud Hole Installed (psia)

(inches) (inches) (inches)

CNP Unit 1 54 7.0 172.5 16.25 2500 CNP Unit 2 54 7.0 172.5 16.25 2500 Range for 16 Units 54 - 60 6.0 - 7.0 157 - 173 15 - 16 2500 Considered Bounding Values 54 6.0 173 16 2500 Used in Analysis The analytical model is shown in Figures 2 and 3 at the end of this Enclosure. The loads considered in the analysis consisted of:

• A design pressure of 2500 pounds per square inch , absolute (psia) at a temperature of 600° Fahrenheit (F) was appl ied to the internal surfaces exposed to internal pressure.

• Bolt/stud preload - The preload on the bounding geometry is calculated as:

c, p

• ID 2 1.1

• 250 0

• 173 2 Ppre l oad = - -- -.- - = 42,338 ps i S*D 2 54 6 where :

Ppreload = Preload pressure to be applied on modeled bolt (psi) p = Internal pressure (psi)

ID = Largest inside diameter of RPV (inches)

C = Bolt-up contingencies (+10%)

s = Least number of studs D = Smallest stud diameter (inches)

• 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 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.

Enclosure to AEP-NRC-2018-33 Page4 Flaw Tolerance Evaluation A flaw tolerance evaluation was performed and documented in Reference 2 using the results of the stress analysis to determine how long it would take an initial postulated flaw to reach the ASME Section XI (Reference 1) allowable flaw size. A linear elastic fracture mechanics (LEFM) evaluation consistent with Subsection IWB-3500 of Reference 1 was performed.

At four flaw depths of a 360° inside-surface-connected , partial-through-wall circumferential flaw, stress intensity factors (Ks) 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 calculation . The circumferential flaw is modeled to start between the 1oth and 11th flange threads from the top end of the flange 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 of this Enclosure, for the flaw model with alt= 0.77 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 below 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 . alt profile.

Table 2: Maximum K vs. alt Kat Crack Depth (ksi~in)

Load 0.02 alt 0.29 alt 0.55 alt 0.77 alt Preload 11 .2 17.4 15.5 13.9 Preload + Heatup +

13.0 19.8 16.1 16.3 Pressure Because a postulated flaw is considered in this evaluation , a conservative LEFM approach consistent with Appendix G of Reference 1 is used to determine the allowable flaw size.

Appendix G of Reference 1 applies a structural factor of 2 to the membrane stress and a structural factor of 1 to the thermal stress. In this evaluation, the conservative factor of 2 will be applied to all stresses. The acceptance criterion based on allowable stress intensity factor is:

K1 < K1cf2 for normal operating condition Where K1 = Applied stress intensity factor (ksi~in)

K1c = Lower bound fracture toughness

Enclosure to AEP-NRC-2018-33 Page 5 The fracture toughness K1e for normal operating temperature is obtained from Figure G-2210-1 of Appendix G of ASME Section XI (Reference 1) for a material operating in the upper shelf region (normal operating temperature). The flaw tolerance evaluation in the EPRI report (Reference 2) used the value of K1e= 220 ksi"in , which is the maximum value allowed for the applicable conditions. Therefore , K1el2 results in an allowable K value of 110 ksi"in. As can be seen from Table 2 above, the allowable K is not exceeded for all crack depths up to the deepest analyzed flaw of alt = 0.77. Accordingly, the allowable flaw depth of the 360° circumferential flaw is at least 77% of the thickness of the flange during normal operation.

However, because most of the applied K in Table 2 above is due to preload that occurs at a lower temperature, the allowable K at preload temperature must also be checked. The fracture toughness K1e during preload is based on the reference temperature for nil ductility transition (RT ndt) of the vessel flange materials and the assumed flange temperature at the time of preload . Appendix B of Reference 2 contains information from 28 nuclear plants, including flange temperature during bolt preload (T) and RTndt values . Using the smallest difference between T and RTndt, (T - RTndt = 0°F) , and the equations in Figure G-2210-1 of Appendix G of Reference 1, the corresponding value of K1e is 53 .9 ksi"in . For the postulated flaw considered in the analysis, using a structural factor of 2 brings the allowable K to 27 .5 ksi"in . Therefore , the allowable flaw depth of the 360° circumferential flaw is at least 77% of the thickness of the flange , even during bolt-up.

The RTndt of the vessel flange region at CNP Unit 1 is 28°F and at Unit 2 is 30°F. The flange temperature during bolt preload for both units is assumed as the procedural minimum of 60°F.

Thus , flange T - RTndt for CNP Unit 1 and Unit 2 are 32°F and 30°F, respectively, which are bounded by the value of 0°F used in Reference 2.

For the crack growth evaluation, an initial postulated flaw size of 0.2 inches was chosen consistent with the flaw acceptance standards in Subsection IWB-3500 of ASME Section XI (Reference 1). The deepest flaw analyzed is alt= 0.77 because of the limits of the model. Two load cases are considered for fatigue crack growth : heat-uplcooldown and bolt preload. The heat-uplcooldown 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 load case is assumed to be present and constant during the load cycling of the heat-uplcooldown load case .

The bolt preload load case is conservatively assumed to occur five times per year, and does not include thermal or internal pressure. The resulting crack growth was determined to be negligible (0.005 inches over 80 years of operation) 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) .

Enclosure to AEP-NRC-2018-33 Page 6 The bounding stress analysis / flaw tolerance evaluation presented above shows that the th read in flange component at CNP Units 1 and 2 is very flaw tolerant and can operate for 80 years without violating ASME Section XI (Reference 1) safety margins. This demonstrates that the thread in flange examinations can be eliminated without affecting the safety of the RPV.

Operating Experience Review Summary The EPRI report (Reference 2) concludes that the examination 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 3 below (reproduced from Table 3-1 of Reference 2) , not a single unit has reported detecting a reportable indication in more than 10,000 examinations conducted . The 94 units identified in Table 3 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 includes 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 nuclear steam supply system designs (i.e. Babcock & W ilcox, Combustion Engineering , and Westinghouse) .

Table 3: Summary of Survey Results - U. S. 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, Reference 2 discusses studies conducted in response to the issuance of the Anticipated Transient Without Scram (ATWS) Ru le (Reference 4) by the NRC . This rule was issued to requi re design changes to reduce expected A TWS frequency and consequences . Many studies have been conducted to understand the ATWS phenomena and key contributors to successful response to an A TWS event. Reference 2 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 (Reference 5) for PWRs, the ASME Service Level C pressure of 3200 pounds per square inch gauge 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.

Add itionally, there was the concern that steam generator tubes might fail before other RCS components , with a resultant bypass of containment. Key find ings from these studies are that

~

I Enclosure to AEP-NRC-2018-33 Page 7 the RPV flange ligament was not identified as a weak link and that other RCS components were significantly more limiting . Thus, there is substantial structural margin associated with the RPV flange .

In summary, Reference 2 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, l&M proposes that this request, including the EPRI report (Reference 2), provides an acceptable technical basis for eliminating the requirement fo r this examination because the alternative maintains an acceptable level of quality and safety.

The EPRI report provides the basis for elimination of the RPV threads in flange examination requirement fou nd in Reference 1 (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 , refueling outage critical path time for these examinations, and additional time at reduced RCS water inventory.

For these reasons, l&M 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 remainder of the fourth 10-year ISi interval , which is scheduled to end on February 29, 2020 .

7. Precedent The NRC has authorized the use of an alternative examination of the RPV threads in flange for several operating nuclear plants. Table 9-1 of the EPRI report (Reference 2) identifies 28 units that submitted alternative requests based on an earlier version of that report, EPRI Report 3002007626 (Agencywide Documents Access and Management System (ADAMS)

Accession Number ML16221A068). In addition , the NRC issued a Safety Evaluation to Duke Energy authorizing their proposed alternative request for nine units on December 26 , 2017 (Reference 6) , incl uding Brunswick Steam Electric Plant, Unit 1, Catawba Nuclear Station ,

Unit 2, Shearon Harris Nuclear Power Plant, Unit 1, McGuire Nuclear Station , Units 1 and 2, Oconee Nuclear Station , Units 1, 2, and 3, and H .B. Robinson Steam Electric Plant, Unit 2.

Enclosure to AEP-NRC-2018-33 Page 8

8. References

1. ASME Boiler and Pressure Vessel Code,Section XI , 2004 Edition , no Addenda .

2. Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements. EPRI ,

Palo Alto , CA: 2017. 3002010354, Final Report, December 2017 .

3. 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.

4. 10CFR50.62, Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants. Federal Register, Vol. 49, No. 124, June 26, 1984.

5. SECY-83-293, "Amendments to 10CFR50 Related to Anticipated Transients Without Scram (ATWS) Events," U.S. Nuclear Regulatory Commission , Washington , D.C.,

July 19, 1983.

6. Brunswick Steam Electric Plant, Unit No. 1; Catawba Nuclear Station , Unit No. 2; Shearon Harris Nuclear Power Plant, Unit 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 lnservice 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).

Enclosure to AEP-NRC-2018-33 Page 9 R86.5" 8.5" 12.0" 17.0" 7.0" 16.0" R83.75" R4.5" II R85.69" Figure 1 - Modeled Dimensions

Enclosure to AEP-NRC-2018-33 Page 10 Figure 2 - Finite Element Model Showing Bolt and Flange Connection

Enclosure to AEP-NRC-2018-33 Page 11 Figure 3 - Finite Element Model Mesh with Detail at Thread Location

Enclosure to AEP-NRC-2018-33 Page 12 Figure 4 - Cross Section of Circumferential Flaw with Crack Tip Element Inserted After 10th Thread from Top of Flange