ML20207C120
| ML20207C120 | |
| Person / Time | |
|---|---|
| Issue date: | 05/17/1999 |
| From: | Samson Lee NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| PROJECT-690, TASK-***, TASK-OR NUDOCS 9906020203 | |
| Download: ML20207C120 (60) | |
Text
{{#Wiki_filter:. $o cro o k UNITED STATES i j NUCLEAR REGULATORY COMMISSION f$ WASHINGTON, D.C. 20666-0001
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C May 17,1999 ORGANIZATION: Nuclear Energy institute (NEI)
SUBJECT:
SUMMARY
OF MEETING WITH NEl ON METAL FATIGUE On April 27,1999, representatives of NEl roet with the Nuclear Regulatory Commission (NRC) staff in Rockville, Maryland, to discuss the industry's evaluation of fatigue effects for license renewal. A list of meeting attendees is presented in Attachment 1. The NEl meeting handout is presented in Attachment 2. The staff's meeting handout is presented in Attachment 3. At the meeting, NEl also provided the staff with a revised discussion on the " weighted average" approach for fatigue evaluations (Attachment 4). The staff has identified Generic Safety issue 190 (GSI-190)," Fatigue Evaluation of Metal Components for 60-Year Plant Life," to address the effects of water environment on the fatigue resistance of metal components for license renewal. NEl submitted several Electric Power Research institute (EPRI) reports on the evaluation of metal fatigue for license renewal indicating that GSI-190 is not a concern. One of these EPRI reports is specific to Calvert Cliffs and is referenced by the Baltimore Gas and Electric Company (BGE) in its license renewal application for Calvert Cliffs. The staff reviewed the Calvert Cliffs specific report and issued a request for adGitional information (RAl). After discussions with BGE, the staff decided to redirect the RAI to NEl because they involve generic information regarding EPRI reports submitted by the industry. By letter dated April 8,1999, NEl responded to the staff's RAl. By letter dated April 20,1999, NEl supplemented its RAI response. The staff discussed the NEl response at the meeting. The following three main issues are addressed in the steff's RAl: (1) carbon steel component fatigue evaluation using a " strain threshold," (2) stainless steel component fatigue evaluation considering the recent test data, and (3) the use of a " weighted average" approach in fatigue evaluations. Regarding the " strain threshold," the staff indicated that the value used by the Argonne National Laboratory (ANL)is based on the mean of the data with a margin of 1.7. Regarding the stainless steel component fatigue evaluation considering the recent ANL test data, the staff indicated that the fatigue / environment adjustment factor should be based on a ratio of fatigue life in an air environment at room temperature and in a water environment at operating temperature. The staff indicated that the existing design code factors that could be attributable to the effects of moderate environment should not exceed 3 for carbon / low-alloy steel and 1.5 for stainless steel. Regarding the use of a " weighted average," NEl indicated that this approach is used to remove excessive conservatism in calculating strain rate due to " dead time" between transients. NEl indicated that the EPRI reports were based on the maximum temperature and maximum oxygen level in the " weighted average" calculations. qlfdf'] [ D. l' C M133 W
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V -2 May 17, 1999 As a result of the meeting, it is agreed that: (1) carbon steel components could be acceptable if the industry uses the threshold suggested by the staff, (2) stainless steel components would be further evaluated by the industry to address the recent data, and (3) the industry's " weighted average" approach may be acceptable pending further staff review. The participants also discussed briefly plant-specific options for BGE to address GSI-190 should the schedule of the NEl effort or the staff's generic resolution of GSI-190 does not support the BGE licensing decision. The Office of Nuclear Regulatory Research (RES) is developing a generic resolution to GSI-190. RES is evaluating the increase in pipe failure frequency under design basis transients when the effects of water coolant environment are considered. The staff expects to complete the resolution of this issue by December,1999. / <h am Lee, Sr. Materials Engineer Engineering Section License Renewal and Standardization Branch Division of Regulatory improvement Programs Project No. 690 Attachments: As stated cc: w/ Attachments: See next page l Distribution: See next page DOCUMENT NAME:G:\\ WORKING \\ LEE \\ SUM _427,WPD MM/ OFFICE LA f/ RLSB RLSB:S@ EMEB:BC DE:DD. G RLSB:BC m_ NAME E$ piton SLee PTKuo 1 Elmbro #E-RWe5fgbh/ CGrime([,d DATE 6 /'/ /99 Y/ P/99 i /10 /9 9 'f//e /99 5 /$ /99 6 / 11/99 I OFFICIAL RECORD COPY
ATTENDANCE LIST NRC MEETING WITH NUCLEAR ENERGY INSTITUTE ON METAL FATIGUE APRIL 27,1999 NAME ORGANIZATION Sam Lee NRC/NRR/ DRIP /RLSB John Fair NRC/NRR/DE Keith Wichman NRR/EMCB Mark Hartzman NRR/EMEB Kamal Manoly NRR/EMEB Frank Chemy RES/DET Jit Vora RES/DET/EMMEB Ed Hackett RES/DET/EMMEB Joe Muscara RES/DET/MEB Omesh Chopra Argonne National Lab. Dick Wessman NRR/DE Chris Grimes NRC/RLSB P.T.Kuo NRC/RLSB Jenny Weil McGraw-Hill Alice Carson Bechtel Power Corp. Gary L. Stevens StructuralIntegrity Associates Paul D. Manbeck BGE John Carey EPRI Robert E. Nickell EPRI consultant Har Mehta GE Nuclear Energy Todd Conner Baltimore Gas & Electric B. W. Doroshuk BGE i Doug Walters NEl l E. Imbro NRR/EMEB J. Strosnider NRR/DE i J I
5 Distribution: Hard cooy PUBLIC Docket File l RLSB RF N. Dudley, ACRS - T2E26 E. Hylton E-mail: R. Zimmerman W. Kane D. Matthews S. Newberry C. Grimes C. Carpenter J. Strosnider R. Wessman E. Imbro W. Bateman J. Calvo i H. Brammer I T. Hiltz G. Holahan T. Collins C. Gratton B. Boger R. Correia R. Latta J. Moore J Rutberg R. Weisman M. Zobler J. Craig M. Mayfield S. Bahadur A. Murphy W. McDowell S. Droggitis RLSB Staff G. Tracy A. Thadani C. Julian J. Fair K. Wichman F. Chemy M. Hartzman K. Manoly E. Hackett J. Muscara
NUCLEAR ENERGY INSTITUTE (License Renewal Steering Committee) Project No. 690 cc: Mr. Dennis Harrison Mr. Robert Gill U.S. Department of Energy Duke Energy Corporation NE-42 Mail Stop EC-12R Washington, D.C. 20585 P.O. Box 1006 Charlotte, NC 28201-1006 Mr. Ricard P. Sedano, Commissioner Mr. Charles R. Pierce State Liaison Officer Southern Nuclear Operating Co. Department of Public Service 40 inverness Center Parkway 112 State Street BIN B064 Drawer 20 Birmingham, AL 35242 Montipelier, Vermont 05620-2601 Mr. Douglas J. Walters Mr. Barth Doroshuk Nuclear Energy Institute Baltimore Gas & Electric Company 1776 i Street, N.W. 1650 Calvert Cliffs Parkway Washington, DC 20006 Lusby, Maryland 20657-47027 DJW@NEl.ORG National Whistleblower Center Chattooga River Watershed Coalition 3233 P Street, N.W. P. O. Box 2006 Washington, DC 20007 Clayton, GA 30525 Mr. William H. Mackay Mr. John Carey Entergy Operations, Inc. Electric Power Research Institute Arkansas Nuclear One 3412 Hillview Avenue 1448 SR 333 GSB-2E Post Office Box 10412 Russellville, Arkansas 72802 Palo Alto, CA 94303
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April 11,1999 TO: Dr. John J. Carey FROM: Dr. Robert E. Nickell, Applie'd Science & Technology Modifled by Gary L Stevens, StructuralIntegrity Associates, 4/26/99
SUBJECT:
Discussion of the F. Time / Stress-Weighted Average Approach i Several EPRI projects have applied an explicit reactor water environmental. effects ] methodology [1] to selected fatigue-sensitive component locations in an older vintage .l Combustion Engineering PWR plant [2], an older vintage Westinghouse PWR plant [3], ) and two newer vintage GE BWRs [4,5]. In three of the four applications [2,3, and 4], actual transient histories obtained through a Fatigue Management System (FMS) were used in lieu of the idealized design-basis transients. One of the physical parameters required for the reactor water environmental effects calculations is strain rate, which may vary substantially at a component location during an actual plant transient, but which can be approximated as a constant value for idealized design-basis transients. 4 4 Figure 1 shows the difficulty that can be encountered when determining strain rates for actual transient data. In order to account for the variation in strain rate during actual j transients, References [2,3, and 4) relied on a Time / Stress-Weighted Average Approach for the environmental shift factor between two extreme points in a transient, calculated from: F fv = (1/W) f (a, - o,.i) F.{c',S,DO,T} dt (1) where F.,yis the environmental shift factor obtained using the Time / Stress-Weighted Average Approach between the low point (valley) in a transient and the next high point (peak) in that transient; o, and o,_i are the stress values at two successive transient history sampling points between the valley and peak; F. is the environmental shift factor, as a function'of strain rate (c'), dissolved oxygen content (DO), temperature (T), and material sulfur content (S) between the two successive transient sampling points; and W is the ' time-weighted stress change, given by: W = j (o, - e,.i) dt (2) 1
L l The instantaneous strain rate, c' (%/sec), is calculated from: l 100(o, - e,_i)
- g. =
Eat where E is the elastic modulus and at is the time difference in seconds between the two L stress sampling points in the transient history. . Questions have been raised about the conservatism and accuracy of the Time / Stress-Weighted Average Approach environmental shift factor. Therefore, the following verification exercises are provided in an attempt to respond to those questions. First, consider the case of a cyclic stress-time history represented by a saw-tooth pattern, with a positive stress amplitude peak ofo, and a negative stress amplitude valley of-o., as shown in Figure 1. For a com-tant increasing stress from the valley to the peak (i.e., a straight line), the strain rate during the entire period ofincreasing stress is a constant. Then, the strain rate given by Equation (3) is as follows:
- g., 100[a,-(o )],
- 200o, o
E(tg - t.ii,y) E(tg - tv ii,y) y Note also that the combination of Equations (1) and (2) for the case of constant strain - rate, with all other variables remaining the same, gives the following: F, = F {s') In other words, for a constant strain rate between a valley and the next peak, the F, calculated with the Time / Stress-Weighted Average Approach is identical to the F, calculated with constant strain rate (the " Standard" Approach). Second,' consider a variation on the caw-tooth stress history, as shown in Figure 3. For the first portion ("A") of the stress history that increases from -a, to o,/2, the following is true: r I o (t) = - o, + (t/t) e, for 0 s t s 3/2r L 2 4
The second portion ("B") of the stress history flattens out, and is given by: o (t) = % o, + [1/(20T)] (t - 3/2t)o, for(3/2)r s t s [(3/2)r + 10t] For this case, W is calculated as (9/4)c t, and W, is calculated as So t, so that W = W^ 4 o o + W = (29/4) e,t. Then: F.,y = (9/29)F {100c/(ET)) -(20/29)F,{5c/(ET)} i In other words, the F, calculated from the Time / Stress-Weighted Average Approach between a valley and the next peak for this case consists of about a 30% weighting for the higher strain rate, and a 70% weighting for the lower strain rate portion of the saw-tooth. The Standard Approach would be based on a strain rate approximately 3.5 times higher than that for the low strain rate (more conservative) portion ("B") of the stress history. If it is assumed that portion "B" (the lowest strain rate for this case) has the lowest strain rate affected by the methodology for computing F, from Reference [1] (i.e.,0.001%), the computed F values are 3.24 for the Time / Stress-Weighted Average Approach versus 3.06 for the Standard Approach. Therefore, the Time / Stress-Weighted Average Approach is more conservative than the Standard Approach for this case. J Third, consider a second variation on the saw-tooth stress history, as shown in Figure 4. j The first portion ("A") of the stress history assumed for this case is identical to that assumed for Case 2 above. However, the second portion ("B") of the stress history is given by: o (t) = % o, + (10/t)(t - 3/2 t) o, for(3/2)r s t 5 [(3/2)t + (1/20)t] For this case, W is calculated as (9/4)c t, as before, and W, is calculated as (1/40)o,t, 4 o so that W = W + W = (91/40)o,t. Then: 4 F.,y= (90/91)F {100o/(Et)) + (1/91)F {1000o/(ET)} In other words, the F, calculated from the Time / Stress Weighted Average Approach between a valley and the next peak for this case consists almost entirely of the F, calculated at the lower strain rate (conservative) that governs for the portion "A" of the stress increase. The reason for this is the small weighting that accrues to the short time 3
1 l period over which the strain rate is high for the last one-quarter of the stress increase. The Standard Approach would be based on a strain rate approximately 1.29 times higher than that for the low strain rate (more conservative) portion ("A") of the stress history. If it is assumed that portion "A" (the lowest strain rate for this case) has the lowest strain rate affected by the methodology for computing F, from Reference [1] (i.e., 0.001 %), the computed F, values are 3.60 for the Time / Stress-Weighted Average Approach versus 3.49 for the Standard Approach. Therefore, the Time / Stress-Weighted Average Approach is more conservative than the Standard Approach for this case. In summary, the Time / Stress-Weighted Average Approach to F,, calculations is either: (1) identical to the Standard Approach for constant strain rates; or (2) more accurate and, generally, more conservative than the Standard Approach for continuous and completely connected transients. For the case of stress peaks separated far in time, the Standard Approach is more conservative due to the artificially low strain rate calculated on the basis of the time separation between the stress peaks and valleys. j l
References:
[1] EPRI Report No. TR-105759, "An Environmental Factor Approach to Account for Reactor Water Effects in Light Water Reactor Pressure Vessel and Piping Fatigue Evaluations," December 1995. [2] EPRI Report No. TR-107515, " Evaluation of Thermal Fatigue Effects on Systems Requiring Aging Management Review for License Renewal for the Calvert Cliffs Nuclear Power Plant," December 1997. [3] EPRI Report No. TR-110043, " Evaluation of Environmental Fatigue Effects for a j Westinghouse Nuclear Power Plant," April 1998. i [4] EPRI Report No. TR-110356, " Evaluation of Environmental Thermal Fatigue Effects on Selected Components in Boiling Water Reactor," April 1998. [5] EPRI Report No. TR-107943, " Environmental Fatigue Evaluations of Representative BWR Components," June 1998. I e 4
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