ML13051A740
| ML13051A740 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/18/2013 |
| From: | John Stanley Calvert Cliffs, Constellation Energy Group, EDF Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML13051A740 (15) | |
Text
Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 CENG.
a joint venture of 0
Eorgstry*eDF CALVERT CLIFFS NUCLEAR POWER PLANT February 18, 2013 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
SUBJECT:
Document Control Desk Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No. 50-318
Response
to Request for Additional Information Relief RR-ISI-04-07A. Dissimilar Metal Butt Welds Baseline Examinations Request
REFERENCES:
(a)
Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC),
dated June 7, 2012, Relief Request for Unit 2 Dissimilar Metal Butt Welds Baseline Examinations (RR-ISI-04-07A)
(b)
Letter from Ms. N. S. Morgan (NRC) to Mr. G. H. Gellrich (CCNPP),
dated February 4, 2013, Calvert Cliffs Nuclear Power Plant, Unit No. 2 -
Request for Additional Information Regarding Relief Request RR-ISI-04-07A, "Dissimilar Metal Butt Welds Baseline Examinations" (TAC No. ME8871)
In Reference (a), Calvert Cliffs Nuclear Power Plant, LLC submitted a relief request for Calvert Cliffs Unit 2 (RR-ISI-04-07A) for authorization of a proposed alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Code Case N-770-1.
In Reference (b), the Nuclear Regulatory Commission requested additional information to support their review of Reference (a). The responses to Reference (b) are attached.
There are no regulatory commitments contained in this letter.
Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.
Very truly yours, for James J. Stanley Manager-Engineering Services JJS/KLG/bjd
Document Control Desk February 18, 2013 Page 2
Attachment:
(1)
Response to Request for Additional Information - Relief Request for Calvert Cliffs Unit 2, RR-ISI-04-07A cc:
B. Vaidya, NRC W. M. Dean, NRC Resident Inspector, NRC S. Gray, DNR
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2, RR-ISI-04-07A Calvert Cliffs Nuclear Power Plant, LLC February 18, 2013
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A INTRODUCTION Calvert Cliffs Nuclear Power Plant, LLC (Calvert Cliffs) staff and the Nuclear Regulatory Commission (NRC) staff participated in a telephone call on January 23, 2013 to discuss relief request RR-ISI-04-07A.
Several items discussed included determination of the maximum hypothetical flaw size possible in the unexamined region of weld 30RC-21 B-10, determination of a maximum assumed flaw repair depth to be used in weld residual stress calculations and evaluation of service lifetime of the largest hypothetical flaw in weld 30RC-21B-10. As the result of a previous Request for Additional Information response and the discussion during the telephone call, the NRC staff has requested additional information.
RAI QUESTION 1:
Please provide the documentation produced during weld fabrication that indicates there were no flaw repairs in excess of l Opercent through-wall depth on weld 30RC-21B-JO.
CCNPP RESPONSE 1:
In lieu of providing documentation produced in 1971, we have performed a thorough review of the available fabrication records and have determined that there is not an undocumented major repair of weld 30-RC-21B-10. This conclusion was shared with the NRC staff and is repeated below.
Since the January 23, 2013 telephone discussion, we have re-reviewed the Shop Traveler, Report of Inspections and Weld Inspection Records associated with the fabrication of weld 30-RC-21B-10. Weld 30-RC-21B-10 is in Combustion Engineering piping assembly 504-03 for Calvert Cliffs Unit 2 and was fabricated under Shop Traveler X-96085-078BS.
The butter weld for inservice inspection weld 30-RC-21B-10 is shop weld #15-504 and it was radiograph examined and found unsatisfactory (as reported in RAI response 5.d.ii of Reference 1) on November 2, 1971.
Shop Traveler Operations 26A and 26B to grind, repair, weld, and penetrant test butter weld
- 15-504 was performed on November 6, 1971. Then, butter weld #15-504 was re-radiograph examined and found satisfactory on November 10, 1971 under Shop Traveler Operations #26C. (Note that suffix letters after the operation number are contingency repair operations to that step, which are built into the Shop Traveler.)
The stainless steel safe end for weld 30-RC-21B-10 was fit-up to assembly 504-03 on December 2, 1971 to make shop groove weld #7-504 (designated weld number 30-RC-21B-10). The weld-out (to the outer diameter) of safe end weld #7-504 was made along with two other safe end welds on pipe assembly 504-03 between December 2, 1971 and December 5, 1971, at which time the backgroove of safe end weld
- 7-504 was performed. The backgroove was penetrant test examined (before backwelding) and found satisfactory on December 6, 1971 under Shop Traveler Operations #34. Note that the contingency grind-out Shop Traveler Operations #34A, B, and C for this backgroove are marked "void" on the Shop Traveler. This further substantiates that there was no penetrant test indication on the initial backgroove chipping and grinding from the inner diameter that required further chipping and grinding.
The backwelding of safe end weld #7-504 was also completed December 6, 1971. Grinding preparations for penetrant testing and radiography were performed on December 7, 1971 under Shop Traveler Operation
- 36.
Shop Traveler Operations #36A, B, and C (contingency post-penetrant test grind-outs, and re-welding) are also marked "void" on the Shop Traveler. The first radiography of safe end weld #7-504 is signed off as "OK" on December 9, 1971 in the Shop Traveler. The Report of Inspection ticket Record F00103 also confirms the radiography of safe end weld #7-504 was "SATISFACTORY" on December 9, 1971 for Shop Traveler Operation #37. Since there is no suffix letter on the Report of Inspection ticket I
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A for Shop Traveler Operation #37, as well as no rejection notice number, this also indicates this radiograph was not a re-radiograph after a repair.
It is also noted in the Shop Traveler that of the other two safe end welds going through this same process cycle, safe end weld #6-508 also passed the first radiograph, but safe end weld #12-507 was found "UNSATISFACTORY" during the first radiograph. The Shop Traveler reflects the contingency grind-out, penetrant test grind-outs, and weld repair Operations 37A, B, and C were completed only for safe end weld #12-507. Safe end weld #12-507 was re-radiographed and again found "UNSATISFACTORY" on December 11, 1971, at which time Rejection Notice 8163 was opened to further disposition that weld.
The entire pipe assembly was cleaned for final penetrant testing of all non-ferrous surfaces and ultrasonic testing of clad and base metal on December 18, 1971.
From this review of the subject records, it can be concluded with a great deal of confidence that a major repair (greater than 10% of wall) of weld 30-RC-2 IB-10 was not performed. The Shop Traveler, Report of Inspections and Weld Inspection Records, along with the timeline of the shop activities regarding weld 30-RC-21B-10 reflect that a high quality weld was achieved.
The record review also shows that examination results and in-process repairs of any significance were recorded for other welds, and the rejection notice process of complex repairs was rigidly adhered to. The findings from this review match fully with the fabrication history reported in the Calvert Cliffs Alloy 600 Program Plan Technical Bases Document.
Therefore, from this review there is a high confidence that a major repair did not occur on weld 30-RC-21B-10.
Therefore, imposing a 50% depth inner diameter repair residual stress consideration would yield a residual stress profile that is incorrect for this weld.
A 50% inner diameter repair assumption would be inappropriate for an evaluation concerning this ultrasonic test examination limitation.
RAI OUESTION 2:
In the licensee's response to the staff's RAI question 4 concerning determination of service lifetime, the licensee cites the results presented in Figures 6-17 through 6-20 of WCAP-1 7128-NP and Figures 5-2 and 5-3 of MRP-349 to support a service period of 10 years. Figures 6-17 through 6-20 of WCAP-1 7128-NP present service lifetime data only up to a period of 48 months, thus cannot be used to support the 10 year service lifetime. Figure 5-2 of MRP-349 is for a weld without WRS, thus is not applicable. Figure 5-3 of MRP-349 presents flaw tolerance curves for lifetimes up to 120 months for welds with WRS. However, the staff was unable to find sufficient supporting information concerning the assumed weld repair depth, WRS, weld geometry, and loading conditions, either in MRP-349 or the documents referenced therein, to evaluate the calculations performed. The stafffinds that the available technical information is currently insufficient to support acceptance of the baseline examination for up to 7 years. In order to permit the staff to independently confirm the proposed service lifetime, please provide either a complete lifetime calculation for weld 30RC-21B-1O for staff review or the pertinent data supporting Figure 5-3 of MRP-349, including specific references and their associated page numbers, for the following:
- a.
Depth of weld repair used for the WRS calculations
- b.
WRS data, either in graphical or tabular form, or as 4th order polynomial coefficients
- c.
Pipe inside diameter 2
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A d
Pipe wall thickness
- e.
Total axial stress at the weld location, including contribution ofpressure, temperature and external loading f
Global bending stress at the weld location
- g.
If the stress effect of the safe end stainless steel closure weld is credited for the existing nozzle to weld stress state, state the length of the safe end and provide details of its stress effect
- h.
Calculated growth with time of the largest assumed initial flaw (if available)
CCNPP RESPONSE 2:
We are providing the requested information for both the WCAP-17128-NP (Reference 3) and the MRP-349 (Reference 4) supporting calculations. The most appropriate results for weld 30-RC-21B-10 are from the WCAP-17128-NP calculation for a 10% inner diameter repair with no post-weld heat treatment, which are contained in Figure 6-17 of WCAP-17128-NP. See Response 2g below for a revised Figure.
- a.
Depth of weld repair used for the weld residual stress (WRS) calculations The reactor coolant pump suction and discharge nozzle dissimilar metal weld residual stress was calculated in Reference 2 and used in WCAP-17128-NP. There are three cases of weld repairs at the inner diameter surface, 10%, 25%, and 50%. The dimensions of the inner diameter weld repairs are shown in Figures 4-5 through Figure 4-7 of Reference 2 and tabulated in Table 1 below. The evaluation in MRP-349 (Reference 4) used reactor vessel discharge nozzle weld residual stress from MRP-113 (Reference 5), assuming it is similar to the reactor coolant pump dissimilar metal weld due to its similarity in pipe size.
Table 1 - Inner Diameter Weld Repair Depth for Weld Residual Stress Calculation Weld Residual Stress from Reference 2 used in WCAP-Weld Residual Stress from MRP-1 13 17128-NP used in MRP-349 10% through-wall ID repair 0.281 inch 15.2%
25% through-wall ID repair 0.704 inch through-wall 0.35 inch 50% through-wall ID repair 1.400 inch ID repair
- b.
WRS data, either in graphical or tabular form, or as 4th order polynomial coefficients Figure 1 shows the layout and cut locations of the weld residual stress model (inner diameter repair not shown).
Cut I is the reactor coolant pump safe-end to piping dissimilar metal weld. Note that the geometry and materials are generic since this report was developed for use by the pressurized water reactor owner's group. The weld residual stress calculated in Reference 2 and used in the crack growth evaluation in WCAP-17128-NP is illustrated in Figures 2 and 3 below for a 10% inner diameter repair without post-weld heat treatment.
The weld residual stresses from MRP-1 13 used in the flaw evaluation contained in MRP-349 are listed in Table 2 below.
3
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Figure 1: Residual Stress Cut Locations Cut4~~
Cut 2 Ctiu3 Figure 2: Axial Residual Stress through Dissimilar Metal Weld With 10% Weld Repair, No Heat Treat Cut I Axial Stress (10% ID Repair, No Heat Treat)
-.-*-Ambient Conditions -u*-Operating Conditions I U) 50000 40000 30000 20000 10000
-10000
-20000
-30000 0.00 0.20 0.40 0.60 xft Ratio 0.80 1.00 4
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Figure 3: Hoop Residual Stress through Dissimilar Metal Weld With 10% Weld Repair, No Heat Treat Cut I Hoop Stress (10% ID Repair, No Heat Treat)
Ambient Conditions -s--Operating Conditions 60000o 50000 -J
ý 40000
= 3000010 20000 10000 0
i 0.00 0.20 0.40 0.60 0.80 1.00 x/t Ratio 5
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Table 2 - MRP-113 Weld Residual Stresses Used in MRP-349 Flaw Evaluation Through-wall Hoop Stress (ksi)
Axial Stress (ksi)
Distance Ratio, x/t 0
62.49 47.14 0.05 63.03 49.56 0.1 63.57 50.10 0.15 65.72 47.41 0.2 59.53 21.28 0.25 46.87
-2.42 0.3 30.44
-29.36 0.35 10.24
-49.02 0.4 0.27
-56.30 0.45 3.23
-56.03 0.5 7.81
-52.53 0.55 16.97
-45.52 0.6 23.97
-36.09 0.65 34.75
-22.09 0.7 44.18
-5.93 0.75 53.06 15.35 0.8 53.06 27.74 0.85 53.06 32.05 0.9 54.41 34.48 0.95 56.30 35.56 1
60.00 33.13
- c.
Pipe inside diameter The inside diameter used for weld residual stress in Reference 2 and the dissimilar metal weld crack growth evaluation contained in WCAP-17128-NP is 30 inches. The crack growth evaluation performed in MRP-349 also assumed an inside diameter of 30 inches.
- d. Pipe wall thickness The wall thickness for the dissimilar metal weld crack growth evaluation in WCAP-17128-NP was assumed to be 3 inches. The crack growth evaluation performed in MRP-349 also assumed a wall thickness of 3 inches.
- e.
Total axial stress at the weld location, including contribution of pressure, temperature and external loading The total applied axial stresses at the dissimilar metal weld for the crack growth analysis in MRP-349 are shown in Figure 4 below. The weld residual stress from MRP-113 was used. The total applied axial stresses for the crack growth analysis in WCAP-17128-NP are listed in Table 3 below. Note that a hypothetical undetected circumferential flaw for weld 30-RC-21B-10 with a size of a/l = 0.12 and a/t =
0.381 would still be in the compressive stress region as indicated on Table 3.
6
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Figure 4: Total Axial Stress at Dissimilar Metal Weld Used for MRP-349 100 Inlet wo Resid 90Inlet w/ Resd 8O0 80 e-Inlet W/ Resid Poly-fit 70 it a
Inlet w/Repair 60 Inlet w/Repair Poly-fit /
s0 10
-20
-10
-10 e.
-:30
-40
-50
-60 0
01 0.2 0.3 0.4 0.5 0.6 017 0.8 0.9 Distance through wall Thickness Ratio x/t 7
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Table 3 - Total Axial Stress at Dissimilar Metal Weld Used for WCAP-17128-NP 10% ID Repair without Heat Through-wall Distance Ratio, x/t TReatment Heat Treatment (ksi) 0 3.01 0.05 16.35 0.1 20.13 0.15 24.25 0.2 20.53 0.25 0.16 0.3
-6.35 0.35
-6.59 0.4
-4.89 0.45
-4.81 0.5 2.05 0.55 6.23 0.6 8.89 0.65 13.23 0.7 16.92 0.75 22.54 0.8 28.59 0.85 34.50 0.9 39.97 0.95 38.90 1
47.65 f
Global bending stress at the weld location The global bending stress used for crack growth analysis in WCAP-17 128-NP is listed in Table 4 below.
8
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Table 4 - Global Bending Stress Used in WCAP-17128-NP x/t Global Bending Stress (ksi) 0 8.617 0.05 8.703 0.1 8.789 0.15 8.875 0.2 8.961 0.25 9.048 0.3 9.134 0.35 9.220 0.4 9.306 0.45 9.392 0.5 9.478 0.55 9.565 0.6 9.651 0.65 9.737 0.7 9.823 0.75 9.909 0.8 9.995 0.85 10.082 0.9 10.168 0.95 10.254 1
10.340
- g. If the stress effect of the safe end stainless steel closure weld is credited for the existing nozzle to weld stress state, state the length of the safe end and provide details of its stress effect.
The weld residual stress model dimensions for the stainless steel weld and the dissimilar metal weld are illustrated in Figure 5 below. The center-to-center between the welds is 5.125 inches. The safe-end length is 4.486 inches. The weld residual stress calculations in Reference 2 include all the fabrication procedures and the postulated inner diameter repairs. The stainless steel closure weld is included in the weld residual stress calculation used to generate Figures 6-17 through 6-20 of WCAP 17128-NP. The computation represents that the stainless steel weld was performed last, following the dissimilar metal weld fabrication process and inner diameter repair. Therefore, the crack growth evaluation in Figures 6-7 through 6-20 in WCAP-17 128-NP includes the effect of the stainless steel weld.
9
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Figure 5: Westinghouse Weld Residual Stress Model Dimensions TV. j L
Radial Dimensions I
t I
- h.
Calculated growth with time of the largest assumed initial flaw (if available)
MRP-349 extended the crack growth evaluation period to 10 years using the weld residual stress from MRP-113.
The largest acceptable assumed initial flaws (aj) are listed in Table 5 below. The crack growth vs. time from MRP-349 is illustrated in Figure 6 below. Additionally, the crack growth data that produced Figure 6-17 in WCAP-17128-NP was used to produce the allowable initial flaw sizes for 72 and 84 months (or 6 and 7 years). The resulting maximum acceptable initial circumferential flaw for a 10%
inner diameter repair with no post-weld heat treatment is illustrated in Figure 7 below.
Table 5 - Maximum Allowable Initial Flaw/Wall Thickness (aj/t) for 10 Years MRP-1 13 weld No weld residual R-13wl N dresia residual stress with ID Repair AR = 2 0.482 0.735 AR=3 0.380 0.712 AR = 6 0.259 0.609 AR= 10 0.199 0.396 Note: AR = aspect ratio = flaw length/flaw depth (I/a) 10
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Figure 6: Reactor Coolant Pump Dissimilar Metal Weld Circumferential Flaw Primary Water Stress Corrosion Cracking Growth vs. Time 1.0 0.9 AR2 NoWR AR3 NoWR 0.8 AR6 No*R ARI0 NoWl 75%Aowab 0.7 -
AR2wire O
AR3 w/ repa 0
ARO w/ rope 0
0.6 AR1O w/rop 0.5 0A 0.3 0.0 0
2 4
6 8
10 12 14 16 18 20 22 24 Time (years)
Circumferential Flaw PWSCC Crack Growth - RCP Inlet/Outlet 11
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - RELIEF REQUEST FOR CALVERT CLIFFS UNIT 2 RR-ISI-04-07A Figure 7: Maximum Acceptable Initial Circumferential Flaws, Accounting for Primary Water Stress Corrosion Cracking and Fatigue Crack Growth, with a 10% Inner Diameter Weld Repair, with No Post-Weld Heat Treatment 0.9 0.8 0.7 I0.5 0.4Time (month) to Reeh 0.43ME Allowable Crack Depth-I 0.2 0.1 0
0 0.1 0.2 0.3 0.4 0.5 Crack DepMi I Lenrgt Ratio. aW
REFERENCES:
- 1. Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC), dated January 10, 2013, Response to Request for Additional Information - Relief Request RR-ISI-04-07A, Dissimilar Metal Butt Welds Baseline Examinations
- 2. Westinghouse Calculation Note, CN-MRCDA-10-1 1, Revision 1, "Finite Element Residual Stress Analysis for a Typical CE-Designed Reactor Coolant Pump Safe-End to Pipe Weld Region,"
October 25, 2012
- 3. Westinghouse Report, WCAP-17128-NP, Revision 1, "Flaw Evaluation of CE Design RCP Suction and Discharge Nozzle Dissimilar Metal Welds, Phase III Study," May 28, 2010
- 4. EPRI Technical Report, Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349), EPRI, Palo Alto, CA:
2012. 1025852
- 5. EPRI Technical Report, Materials Reliability Program:
Alloy 82/182 Pipe Butt Weld Safety Assessment for U.S. PWR Plant Designs (MRP-1 13), EPRI, Palo Alto, CA: 2005. 1009549 12