ML17229B617

Diablo Canyon, Unit 1 - Flaw Evaluation of Residual Heat Removal Suction Weld Joint

ML17229B617
Person / Time
Site:	Diablo Canyon
Issue date:	08/17/2017
From:	Welsch J M Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-17 -070 1700609.40l.R1
Download: ML17229B617 (25)

Text

{{#Wiki_filter:Pacific Gas and Electric Company August 17, 2017 PG&E Letter DCL-17 -070 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 James M. Welsch Vice President, Nuclear Generation Flaw Evaluation of Unit 1 Residual Heat Removal Suction Weld Joint Diablo Canyon Power Plant Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 805.545.3242 Internal: 691.3242 Fax: 805.545.4884

References:

1. PG&E Letter DCL-17-049, "Flaw Evaluation of Unit 2 Residual Heat Removal Suction Weld Joint," dated May 18, 2017 2. PG&E Letter DCL-17 -048, "Request for Approval of an Alternative to the American Society of Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirements for Class 1 and 2 Piping Welds," dated May 18, 2017

Dear Commissioners and Staff:

In accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section XI, 2007 Edition through 2008 Addenda, Paragraphs IWB-3132.3, IWB-3134(b), and IWB-3640, Pacific Gas & Electric Company (PG&E) is submitting the results of the analytical evaluation of a flaw found in Diablo Canyon Power Plant Unit 1, on the 14-inch diameter ASME Code Class 1 weld identified as WIB-228. IWB-3134(b) states, "Analytical evaluation of examination results as required by IWB-3132.3 shall be submitted to the regulatory authority having jurisdiction at the plant site." PG&E identified a circumferential indication in the Unit 1 Residual Heat Removal (RHR) suction pipe at weld joint WIB-228. The indication was found using ultrasound examination techniques during an extent of condition inservice inspection performed during the Unit 1 twentieth refueling outage. WIB-228 is the pipe-to-45 degree elbow weld on the wrought stainless steel RHR hot leg suction pipe inside the containment building downstream of the Loop 4 reactor coolant system hot leg branch connection. The weld/pipe location in Unit 1 is similar to the weld/pipe location in Unit 2, where a weld flaw was found (Reference 1 ). The flaw exceeds the acceptance standards of ASME BPVC Section XI, 2007 Edition through 2008 Addenda, Table IWB-3514-2. The flaw was initially dispositioned as unacceptable. Subsequently, an analytical evaluation was performed based on the rules of ASME BPVC Section XI, Paragraph IWB-3640. A member of the STARS Alliance Callaway

• Diablo Canyon
• Palo Verde
• Wolf Creek Document Control Desk August17,2017 Page 2 PG&E Letter DCL-17-070 The evaluation concluded that the weld containing the flaw is acceptable for continued service for Cycle 21 operation, which will not exceed two years. The analytical evaluation that was performed for Unit 2 (Reference 1) was replicated using the flaw parameters, pipe dimensions, and loads that are applicable to Unit 1. The specific mechanisms considered were thermal shock from cyclic swirl penetration and cyclic thermal stratification, and stress corrosion cracking (SCC). Available pipe surface temperatures from Unit 2 thermocouple monitoring data were used to perform the analytical evaluation for Unit 1, since the configuration of the pipe/weld, materials, operation, and load history are similar for both Units 1 and 2. Vibration was ruled out as a possible cause based on the inspections performed by PG&E, which concluded that the physical evidence indicative of excessive vibration was not present. Because the evaluation could neither rule out nor specify the specific flaw growth mechanism, a conservative analysis was performed combining the effects of two different flaw growth mechanisms: fatigue crack growth and SCC. The results of this conservative analysis were used to justify continued operation of Unit 1 for an additional cycle. However, PG&E is not attributing SCC with respect to risk-informed inservice inspection (RI-ISI) weld inspections requested by Reference 2 due to uncertainty regarding the flaw growth mechanism. The results of the analysis show that it would take 35 months for the as-found flaw depth of 0.2 inch to reach the allowable flaw size using the most limiting crack growth rate. Therefore, the continued operation of Unit 1 for an additional cycle was determined to be acceptable. The analysis, methodology, and results are provided in the Enclosure. PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this letter. If you have any questions or require additional information, please contact Mr. Hossein Hamzehee at (805) 545-4720. Sincerely, Welsch Vice President, Nuclear Generation rntt/4231/50915871-26 Enclosure A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
• Diablo Canyon
• Palo Verde
• Wolf Creek Document Control Desk August17,2017 PG&E Letter DCL-17 -070 Page 3 cc: Diablo Distribution cc/enc: Scott A. Morris, NRC Region IV Administrator (Acting) Christopher W. Newport, NRC Senior Resident Inspector Gonzalo L. Perez, Branch Chief, California Department of Public Health Balwant K. Singal, NRC Senior Project Manager State of California, Pressure Vessel Unit A member of the STARS Alliance Callaway
• Diablo Canyon
• Palo Verde
• Wolf Creek Diablo Canyon Power Plant, Unit 1 Enclosure PG&E Letter DCL-17-070 Residual Heat Removal (RHR) Piping Flaw Evaluation Report 1 e Structural Integrity Associates, Inc. August 04, 2017 Report No. 1700609.401.R1 Quality Program: [:8:1 Nuclear D Commercial. Mr. Mark Sharp Pacific Gas and Electric Company Diablo Canyon Power Plant PO Box 56 Avila Beach, CA 93424 5215 Hellyer Ave. Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.structint.com aalleshwaram@structint.com

Subject:

Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation

Dear Mark:

In response to your request dated 5/1117, this report documents the flaw evaluation of the recently discovered flaw in the RHR Suction Weld Joint at Diablo Canyon Power Plant (DCPP) Unit 1. BACKGROUND During the scheduled refueling outage 1R20 work, ultrasound (UT) inspection on Unit 1 weld WIB-228 identified an ultrasonic reflector at approximately 01:00 on 4-28-17 [15]. This is line 109 of the RHR system "Hot Leg Recirc Before Isolation Valve V-8702." This weld is a to-450 elbow off the Loop 4 hot leg (see Figure 1). This is the same weld/pipe location in Unit 2 (WIB-245) where a weld flaw was found in 2R19. The objective of this report is to perform a flaw evaluation for the flaw found in Unit 1 weld WIB-228 using two flaw growth mechanisms of fatigue and stress corrosion cracking (SCC) to justify continued operation for an additional cycle. Since the flaw was found at a similar location as Unit 2, the flaw evaluation performed for Unit 2 is replicated using the flaw parameters, pipe dimensions, and loads applicable to Unit 1. Based on the conservative treatment of loading uncertainties in the analyses performed for Unit 2 [1], SI recommended the application of temporary thermocouples to both welds of the 45° elbow to better understand the real loading. The temperature monitoring conducted by DCPP for Unit 2 -----------------Toll-Free 877-474-7693 _______________ _ Denver, CO 3D3*792*00n Al<ron, OH 330-899-9753 Mystic, CT 860*536*3982 Albuquerque, NM 505-H72-0123 Poughkeepsie, NY 845*454*61 00 Austin, TX 512-533-9191 San Diego, CA 858*455*6350 Charlotte, NC 71)4-597 -5554 San Jose, CA 408-978*8200 Chattanooga, TN 423-553-1180 Slate College, PA 814*954*n76 Chicago, IL 815-648-2519 Toronto, Canada 905*829*9817 Mr. Mark Sharp August 04, 2017 Report No. 1700609.401.R1 Page 2 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation was implemented by attaching thermocouples to the outside surface of the piping at several locations as shown in Figure 2. A revised analysis for Unit 2 was performed using the thermocouple data [2] to more accurately calculate the thermal stresses. This data shows that the RHR has cyclic stratification of the unisolable horizontal pipe section. This data allowed for the calculation of more accurate local stresses and global bending loads due to stratification and thermal expansion. Since the configuration of the pipe/weld, materials, operation and load history are similar for both Units 1 and 2, the available Unit 2 thermocouple data is used to perform the flaw evaluation for Unit 1. Note that the cracking found in Unit 1 is not as severe as the cracking found in Unit 2. Vibration was not included since it was ruled out as a possible cause. This report documents the loads used to calculate stresses and stress intensity factors to determine crack growth (from both fatigue and SCC) and the allowable flaw size. Based on the evaluation contained herein, the flawed component is acceptable for continued operation. The flaw is not expected to exceed the ASME Code, Section XI allowable flaw size for the next 35 months. DESIGN INPUT Geometry Pipe outside diameter (OD) = 14 inches [3] Pipe thickness = 1.406 inches [3] Loads The flaw location is shown in Figure 1 and is at Node 503 per Reference [3]. All piping loads are obtained from Reference [3] at Node 503 and are shown in Figure 3. Mr. Mark Sharp August 04, 2017 Report No. 1700609.40l.R1 Page 3 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation (a) Isometric Drawing ilT (b) Piping Model Figure 1. Flaw Location T/C*l.09 llO 11' nu*r I* . I o'H *F (*d11 I r, l * *r *6. o'\1 r 'F ; * *u r .11 I 'J *F {*lO lot f L f *' ** (* ',0. ) Figure 2. RHR Piping Thermocouple Locations for Unit 2 Mr. Mark Sharp August 04,2017 Report No. 1700609.401.R1 Page 4 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation FOR CLASS 1 EOUIPMEN'f E L s T" 5 a c -@ 1 Jo.J.c2 5o'3 HFHl 2HIJ?.56 NODE NO. : 503 P&ID DWG. NO.: 102007 NOZZLE ID: LINE NU!o!.BER S6-109-14 l:'li?ING SYS'J'EH : RHR -RCS lWS'l'E:l*l EQOil?MgNT NANE : PIPE -ELBOI-7 COSAY, COSAZ : -0.265 -0.707 0.656 COSBX, COSBY, COSBZ : 0.265 -0.707 -0.656 COSCX, COSCY1 COSCZ : 0.927 0.000 0.375 LOADS GtVt.N 11'1' [ ] SHE:J,L ( ] NO:<',ZLE: F'L.J:\NGE FOP.CE (LBSl MOt-lENT (FT-LBS) L01'illiHG CONDITION FA f'B FC HA MB DEADI*1EIGHT: j)L, 1'HlmNA : -197, THRNNl 64l. 67 2. 'rHRHN3 1461. 9 8.4 , 'rHRNNS -124. 375. 'PHRJ.!Al BIJU. 9 6 6. 1'HRJ*IAJ 635. TllRHA4 716. THRNAS 133. TIIRl>Lll.6 215 . .SEIS!o!IC-I.NER1'IA: DE 990. ODE 1622. £10SGRI 2220. SEIS!-liC ANCHOR MOTION: DE DDE HOSGRI 1300. 1690. 1686. -187. -1513. -1571. -653. -417. 55. -720. -920. -1153. -99. -.:na. -1635. -1914. ]882. 284!. 5197. 87l. 1152. 114 6. 0. -2572. -1740. -lBS. 110. -L265. -2261. -2.581[. -{:)30. -1011. -1612. -1993. 1330. 1795. 3039. 7515. 9519. 11193. 11!)4. 22737. 23590. 2316'7. 1B332. 1021. 17.562. 23395. 23575, 11600. J 211SIJ. 30459. 31343. 3543. 5890. 6155. 7782. 9693. -1380. -3949. -4267. Hll69. 7875. -3026. -9:35. -3911. -.U*1l. 5973. 6799. -5678. -4862. 1962. 2958. {60*1. 3834. 5161. 4378. -1440. 1118. 1107. 6200, 5400. 515. .988. 6549. 6033. 3948. 280!L 632. -507. 2560. 4021. 6738. 1151. 1463. 1751. 1. Make a simple sketch of the equipment, Lhe nozzle, the local and the global coorclLnate system and north <>n:m., in t;..he bl.vcl< above. List ln local coordinates. FA is along the nozzle centeJ:1ine. Positive F."o is from in.s icle to outs ide of equ *pill Em L, Figure 3. Nozzle Loads at Node 503 Operating Conditions Reference [3] defines the operating pressure conditions for the RHR line which is shown in Figure 4. Mr. Mark Sharp August 04,2017 Report No. 1700609.401.R1 Page 5 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Table 9.7.2 Reactor Coolant System Operating Pressure and TemP-erature Modes Pipe Enveloped Line Spec Conditions INN 2NN 3NN 3All 4NB 63 S6 2510 2243 2254 2504 2504 0 Note 13 650 NIS 547 579 579 NIS 109 S6 2510 2254 2254 2504 2504 50 Note 15 NIS NIS NIS NIS 140 Figure 4. Operating Pressures (Note that the operating pressures for the RHR line are taken for line 1 09.) RHR Suction Line Materials The material of the piping is ASTM A-376 TP 316 austenitic stainless steel [3]. Weld groove was installed using ER308 filler metal with 880 flux (SMA W/SA W process) [1]. For the flaw evaluation, the flow stress of the material is required. Using the design temperature of 650°F [3], the stainless steel material properties are as follows: Yield stress, o-y = 18.5 ksi [ 4] Ultimate tensile stress, o-u = 71.8 ksi [ 4] Flow stress, ay= 45.15 ksi (average of (Jy and (Ju) Flaw Size One circumferential ID surface connected flaw was identified in the weld [3]. The flaw sketch is shown in Figure 5. Flaw Length at ID surface (2c) = 4.8 inches [3] Flaw Depth (a)= 0.2 inch [3] Mr. Mark Sharp August 04, 2017 Report No. 1700609.40l.R1 Page 6 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Pipe Thermal Transient Indication #1 60° L 281 ""'-Flow:=:-.... Lookino Downstream /"' r 4.811 Indication Extent Une 109 *weld IAJJB-228 4so Elbow /)II Figure 5. Flaw Configuration Paye 5 of The thermocouple data from the reactor coolant system (RCS) Loop 4 RHR piping at 100% power for Unit 2 [5] is used to perfonn the evaluation for Unit 1. Mr. Mark Sharp August 04,2017 Report No. 1700609.401.R1 Page 7 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation ALLOW ABLE FLAW SIZE DETERMINATION The characterized flaw length is 4.8 inches with a depth of 0.2 inch [3]. The pipe wall thickness is 1.406 inches [3] resulting in a flaw depth-to-thickness ratio of0.14. Using Table IWB-3514-2 of ASME Code Section XI [6], this flaw does not meet the Acceptance Standards and will require disposition per the flaw evaluation procedures of IWB-3600. The stresses are calculated based on the piping loads from Reference [3]. A description of the loads is given below. After the description, the load combinations for all Service Levels are described.

• Deadweight (DW)-Deadweight of piping system, identified as DL
• Thermal (TH)-12 thermal expansion conditions are identified as THRMN1-6 and THRMA1-6. The bounding thermal load is THRMA6, which is thermal accident load for Service Level D. THRMA6 is conservatively applied for all Service Levels to calculate allowable flaw size.
• SAM DE-Seismic anchor movements from DE condition (similar to OBE)
• SAM DD-Seismic anchor movements from DD condition (similar to SSE)
• SAM HS-Seismic anchor movements from HS condition (faulted seismic condition specific to DCPP)
• Seismic DE-Inertia seismic loads for DE condition
• Seismic DD-Inertia seismic loads from DD condition
• Seismic HS -Inertia seismic loads from HS condition With the definition of the piping loadings, the load combinations for each Service Level are defmed as: Primary Loads Service Level A -D W + Pressure (P) Service Level B -DW + P +DE Service Level C-DW + P + DD Service Level D-DW + P + HS Secondary Loads Service Level A -TH Service Level B -TH + SAM DE Service Level C-TH + SAM DD Service Level D -TH + SAM HS The moment loads are shown in Table 1. For the allowable flaw size evaluation, the maximum pressure for the enveloped condition, 2,510 psi shown in Figure 4, was conservatively used for all Service Levels. The applied stresses for each Service Level are shown in Table 2. The stress due to the thermal stratification, 9.964 ksi provided in Table 4, is also used to determine the allowable flaw size. The applied stresses Mr. Mark Sharp August 04, 2017 Report No. 1700609.40l.R1 Page 8 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation including the stress due to the thermal stratification transient for each Service Level are shown in Table 2. Using the defined loading and following Section XI, Appendix C [6] EPFM guidance appropriate for the subject weld, allowable flaw sizes are calculated using pc-CRACK [7], a fracture mechanics program developed by Structural Integrity Associates under its quality assurance program. The allowable flaw size results are shown in Table 3. Table 1. Moment loading for Stress Calculations DW+SEIS Service Level Moments; ft-lbs Msrss, ft-lbs MA MB MC MABC A 1194 -1380 -1440 2325 B 3836 -3342 -4000 6494 c 4737 -4338 -5461 8446 D 7084 -5984 -8178 12389 THRMA6+SAM Service Level Moments, ft-lbs Msrss, ft-lbs MA MB MC MABC A 31343 -4862 -507 31722 B 37498 -8696 -1658 39064 c 39125 -10023 -1970 41174 D 41036 -9240 -2258 42501 Table 2. Applied Stress Summary for Allowable Flaw Size Calculation Service Pressure O'm O'b O'e (I) (ksi) Level (psia) (ksi) (ksi) Thermal Thermal Total Load Cl) Stratification Thermal A 2510 6.248 0.175 2.386 9.964 12.350 B 2510 6.248 0.489 2.939 9.964 12.903 c 2510 6.248 0.635 3.097 9.964 13.061 D 2510 6.248 0.932 3.197 9.964 13.161 Notes: (1) The maximum membrane plus bending stress due to the thermal stratification transient, 9.964 ksi (see Table 4), is added to the secondary bending stress. (2) Global bending stress due to thermal expansion.

Mr. Mark Sharp August 04,2017 Report No. 1700609.401.Rl Page 9 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Table 3. Allowable Flaw Size Results I/ eire 0.2 0.3 0.4 0.5 0.6 >0.75 0.2 0.3 0.4 0.5 0.6 >0.75 Membrane +Bending Membrane Only Level A 0.75 0.66 0.53 0.47 0.43 0.42 0.75 0.74 0.66 0.57 0.51 0.47 Level B 0.75 0.69 0.55 0.48 0.45 0.44 0.75 0.75 0.72 0.64 0.58 0.53 Level C 0.75 0.70 0.58 0.51 0.47 0.44 0.75 0.75 0.75 0.74 0.70 0.64 Level D 0.75 0.73 0.61 0.53 0.48 0.44 0.75 0.75 0.75 0.75 0.75 0.72 Minimum Allowable 0.75 0.66 0.53 0.47 0.43 0.42 0.75 0.74 0.66 0.57 0.51 0.47 a/t CRACK GROWTH ANALYSIS The crack growth analysis herein is performed using the same methodology used in the crack growth calculation for Unit 2 [8]. Weld Residual Stress Analysis The residual stresses in the weld are required for the crack growth evaluation. A finite element analysis was performed to obtain the weld residual stress distribution at the RHR piping elbow weld joint using ANSYS finite element analysis software [9]. The elbow weld is performed in 8 layers. A total of 23 nuggets is defined for this weld. The weld nuggets are defined based on weld size to obtain appropriate heat penetration. A convection heat transfer coefficient of 5.0 Btu/hr-ft2-°F at 70°F bulk ambient temperature is applied to simulate an air environment at the inside and outside surfaces during the application of the weld process. The heat generation rate for the weld, 28 kJ/in, is applied to the weld process. After the application of the weld, a five-cycle normal operating temperature and pressure load step is appended to obtain the stabilized combined residual stresses at room temperature and also at normal operating conditions (NOC). This load step essentially simulates five heatup and cooldown ramp cycles. The resultant residual stress distribution at room temperature is shown in Figure 6. Mr. Mark Sharp August 04, 2017 Report No. 1700609.40l.R1 Page 10 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation 50 40 y = -93.487x3 + 300.16x2-233.59x + 30.826 30 . 20 Vl Vl +-' 10 Vl c:o ::I "'0 *v; <l) 0::: 0 -10 --Residual Stress ......... Poly. (Residual Stress) -20 -30 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 Distance from Inside Surface (in) Figure 6. Through-wall Axial Residual Stress Distribution Thermal Analysis Since the geometry of the RHR piping in Unit 1 Loop 4 is similar to the one in Unit 2 and the same thermocouple data of the RHR piping in Unit 2 is used in this analysis, the stress results of the thermal finite element analyses for Unit 2 [10] is applied in this calculation. Linearized through-wall stresses as well as detailed through-wall stresses on a linear path through the centerline of the flaw, as shown in Figure 7, are extracted from each analysis case. The linearized membrane-plus bending (Pm+Pb) axial stresses are compared within each analysis case to determine the maximum and minimum axial stress time points. Utilizing linearized Pm+Pb stresses is a common practice to assess the stress results within a thermal transient for fracture mechanics analysis. The axial stress results are summarized in Table 4. The through-wall axial stresses at the corresponding minimum and maximum time points are fit into a third order polynomial. The curve fit coefficients are used in the crack growth calculation. Mr. Mark Sharp August 04, 2017 Report No. 1700609.401.R1 Page 11 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Through-wall stress extraction location is at flaw centerline Figure 7. Through-wall Stress Extraction Location Table 4. Through-wall Axial Stress Summary for Thermal Loads Linearized Stresses (ksi) Polynomial Curve Fit Coefficients (ksi) Load Case Pm+Pb Pm+Pb Quality Co c1 c2 C3 (Inside) (Outside) R2 Case 1 Min -22.602 3.283 -30.977 60.245 -42.612 10.146 0.9999 Max 9.964 -14.044 21.232 -76.776 67.253 -19.579 0.9994 Min -22.609 3.281 -30.984 60.249 -42.612 10.146 0.9999 Case 2 Max 9.958 -14.045 21.225 -76.771 67.252 -19.578 0.9994 Min -23.724 3.161 -32.130 61.101 -42.751 10.175 0.9999 Case 3 Max 8.826 -14.155 20.047 -75.815 67.016 -19.521 0.9994 Min -23.733 3.158 -32.138 61.106 -42.752 10.175 0.9999 Case 4 Max 8.818 -14.158 20.039 -75.810 67.014 -19.520 0.9994 Unit Internal 1.841 1.662 1.867 -0.302 0.264 -0.111 0.9975 Pressure Cases: 1. Hot leg connection at 593°F, zero hot leg thermal anchor movement, derived RHR inside surface temperature history from TC1 02 to TC112 thermocouple data. 2. Hot leg connection at 607.4°F, zero hot leg thermal anchor movement, derived RHR inside surface temperature history from TC1 02 to TC112 thermocouple data. 3. Hot leg connection at 593°F, maximum hot leg thermal anchor movement, derived RHR inside surface temperature history from TC102 to TC112 thermocouple data. 4. Hot leg connection at 607.4°F, maximum hot leg thermal anchor movement, derived RHR inside surface temperature history from TC1 02 to TC112 thermocouple data. Mr. Mark Sharp August 04, 2017 Report No. 1700609.401.R1 Page 12 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Stress Intensity Factors The through-wall axial stress distributions for the internal pressure, thermal piping load, thermal stratification transient load, and weld residual stress at 70°F were input into pc-CRACK [7, 11]. The coefficients of third polynomial curve-fit axial stress distributions were input into CRACK, and using "Crack Model309-Semi-Elliptical Circumferential Crack in Cylinder on the Inside Surface," as shown in Figure 12 of Reference [7], the K values were determined. The stress intensity factor (K) values (as a function of crack depth) for each load cases are shown in Figure 8. The crack model is a semi-elliptical crack in cylinder, with varying crack aspect ratio and an initial flaw depth (a) of 0 .2" and an initial a half crack length (c) of 2.4", corresponding to an aspect ratio, c/a of 12.0. 20 10 0 -10 cic ., *v; ::,.:: -20 --------' ' ' ' K vs. a ' ' ' ......... ........ ........ / ....... ..._ ____ -r-..,., -30 --DW Unit Crack Face Pressure -40 --Residual Stress --Thermal Strat. rnax --Thermal Strat. min -50 0 0.2 0.4 0.6 0.8 1 Crack Depth from Inside Surface (in) Figure 8. Stress Intensity Factors 1.2 Mr. Mark Sharp August 04, 2017 Report No. 1700609.401.R1 Page 13 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Fatigue Crack Growth The FCG laws used in this analysis are discussed in the following subsections. In FCG, the individual terms that constitute nominal Kmax and Kmin for the calculation of M< are sum1narized in the following tabulations. Kmax Kresidual Kdead weight Kpressure Kcrack face pressure Kthermal transient max Kmin Kresidual Kdead weight Kpressure Kcrack face pressure Kthermal transient min The individual K values for nominal Kmax are combined (summed) with all appropriate scale factors applied. Similarly, the individual K values for nominal Kmin are combined (summed) with all appropriate scale factors applied. ,1.K is computed by taking the difference of the resulting summed Kmax and Kmin. Note that Kpressure, Kdead weight, Kthermai pipirig load, and Kresiduai are constant loads during the thermal stratification transient and therefore, do not contribute to the range. However, these K values affect the value of the R-ratio (Kmin/Kmax). The number of cycles is determined from the thermocouple data, which is discussed in Section 4.3 .4 of Reference [8]. Fatigue Crack Growth Law for Austenitic Steels in Air The FCG law for austenitic steels in air is per Subsubarticle C-3210 of the ASME Code, Section XI [6], as shown below, but with a multiplier of 2 [12] to account for a PWR environment. where: da/dN = 2*Co(,1.Kt, units of inch/cycle Co c s T R ,1.K n =C*S = 1 0[-10.009 + 8.12x10-4T -1.13 x10-6T"2 + 1.02 x10-9T"3] = 1. 0 when R :::; 0 = 1.0 + 1.8R when 0 < R:::; 0.79 = -43.35 + 57.97R when 0.79 < R < 1.0 =metal te1nperature, op (taken as the average fluid temperature during the transient) = Kmin/Kmax = R-ratio = Kmax-Kmin (range of applied K), ksiv'in = 3.3 per Section XI, Appendix C [6] Fatigue Crack Growth Law using Code Case N-809 Equation The FCG law for Type 304 and Type 316 stainless steels and associated weld metals from ASME Code Case N-809 [13] is also considered in order to compare the results of the FCG law shown above. \ da/dN = Co-,1.Kn, units of inch/cycle Mr. Mark Sharp August 04, 2017 Report No. 1700609.40l.R1 Page 14 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Flaw Evaluation where: Co n c ST = scaling parameter that accounts for the effect of loading rate and environment on fatigue crack growth rate = c ST SR SENV =slope of the log (da/dN) versus log curve= 2.25 = nominal fatigue crack growth rate constant = 4.43 x 10-7 for 2: = 0 for < = stress intensity factor range, ksi"'in = 1.10 ksi"'in = parameter defining effect of temperature on FCG rate = e-25161TK for 300°F :::; T:::; 650°F = 3 .39x1 05 e[C-25161TK)-0.030ITK] for 70°F:::; T < 300°F = metal temperature, op =parameter defining the effect ofR-ratio on FCG rate = 1.0 for R < 0 = 1 + e8.02(R-0.748) for 0:::; R < 1.0 = Kmin/Kmax = R ratio = parameter defining the environmental effects on FCG rate = TR03 = loading rise time, sec = [(T-32)/1.8+273.15], K Stress Corrosion Crack Growth SCC is a time dependent phenomenon and occurs during sustained loading conditions. Given that the great majority of plant operation is at steady state normal operating conditions (NOC), SCC is defined by the stress condition at NOC. SCC is defined to be active when the K at steady state NOC is a positive number. If SCC is active, crack growth is determined for one-month time period. Alternating one-month blocks (periods) of FCG and SCC growth (if SCC is active) are used to calculate total cumulative crack growth. Even though the stress due to the thermal stratification is cyclic in nature, the frequency is high enough that the maximum thermal stratification stress is treated as a constant stress in the SCC growth evaluation. For SCC, the K is calculated in the following tabulation. K Kresidual Kpressure Kcrack face pressure Kdeadweight Kthermal transient max Mr. Mark Sharp August 04, 2017 Report No. 1700609.40l.R1 Page 15 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation SCC Growth Law The SCC growth law for the Type 304 stainless steel is taken from BWRVIP-14-A [14]. Note that the use of this equation for the evaluation of Type 316 weld is conservative since Type 316 austenitic stainless steel is more resistant to IGSCC. ln(da) = 2.181ln(K)-0.787 Cond-0.586 +0.00362 ECP+ 6730-33.235 ili where: da/dt K Cond ECP SHE TABs crack growth rate (change in crack depth per unit time), mm/s stress intensity, MPa.Vm average conductivity (determined at room temperature), J..LS/cm electrochemical corrosion potential, m V(SHE) Standard Hydrogen Electrode temperature, K The above crack growth equation is converted to following simpler K-dependent form: where, da/dt =crack growth rate, inch/hour C, n = material constants K =stress intensity factor, ksh/in A metal temperature of 450oF is applied to calculate the material constants, C and n. The calculated constants, C = 2.42x10-7 and n = 2.181, are applied to calculate SCC growth. No stress corrosion crack growth threshold is considered. Results of Analysis The crack growth results are shown in Figure 9. The allowable flaw size is calculated based on the 1/circumference at each crack depth since the 1/circumference is also changed for every month as shown in Figure 10. At the weld location, it takes 41 months for the initial flaw of 0.2 inches to reach the allowable flaw size when ASME Code FCG equation is used. It takes 35 months for the initial flaw of 0.2 inches to reach the allowable flaw size when Code Case N-809 FCG equation is used. Table 7 shows the monthly crack growth per FCG and SCC. Mr. Mark Sharp August 04,2017 Report No. 1700609.401.R1 Page 16 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Table 7. Monthly Crack Growth Crack Depth (in) Time ASME Code Eq. (Month) Austenitic Steel in Code Case N-809 Air 0 0.2000 0.2000 FCG 0.2520 0.2982 1 sec 0.3486 0.3706 FCG 0.4028 0.4626 2 sec 0.4305 0.4735 FCG 0.4788 0.5355 3 sec 0.4869 0.5368 FCG 0.5302 0.5796 4 0.5320 0.5796 sec FCG 0.5653 0.5877 5 sec 0.5655 0.5877 FCG 0.5882 0.5917 6 sec 0.5917 0.5882 FCG 0.6075 0.6125 12 sec 0.6075 0.6125 FCG 0.6290 0.6328 24 0.6290 0.6328 sec FCG 0.6358 0.6357 35 sec 0.6358 0.6357 FCG 0.6358 -41 sec 0.6358 - Mr. Mark Sharp August 04, 2017 Report No. 1700609.40l.Rl Page 17 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation 1.1ooo I ,___ __ _ 1.0000 0.9000 0.8000 E -;; 0.7000 b. <lJ 0 0_6000 u 0.5000 0.4000 0.3000 0.2000 u 1U Fatigue + SCC Growth -----ASM E Austenitic Steel in Air X2 + SCC --Code Case N-809 + SCC -----Allowable for ASME 1\utenitic Steel in Air X2 + SCC --Allowable for Code Case N-809 + SCC -----.... ____ _ intercepts at 35 months when the Code Cade N-809 FCG equation is used intercepts at 41 months when the ASME Austenitic Steel in Air X2 FCG equation is used iU 30 40 Time (month) Figure 9. Fatigue plus SCC Growth Results from an Initial Flaw Size of 0.2 in bO Mr. Mark Sharp August 04, 2017 Repmi No. 1700609.401.R1 Page 18 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Q) u c 0.90 0.80 0.70 0.60 J;! , E o.so ::J *u ::::::. 0.40 0.30 0.20 0 10 Crack Length I Circumference Growth -----ASM E Austenitic SteEl in Air X2 + SCC --Code Case N-809 + SCC 20 30 40 Time (month) Figure 10. Crack Length per Circumference Growth Results CONCLUSIONS bU The evaluation herein determines allowable flaw depths and expected crack growth for the RHR suction line at DCPP, Unit 1 where cracking was recently discovered. The cracking is located in the pipe-to-45° elbow weld (WIB-228). Using the methods of ASME Section XI, Appendix C, allowable flaw depth at the crack location was determined as a function of flaw circumference ratio. The evaluation for Unit 2 [2] was performed according to the ASME Section XI 2001 Edition through 2003 Addenda [6]. In this report, the analytical evaluation that was performed for Unit 2 was replicated using the flaw parameters, pipe dimensions, and loads that were applicable to Unit 1, since the configuration of the pipe/weld, materials, operation and load history are similar for both Units 1 and 2. However, the applicable ASME Section XI Code of reference for Unit 1 is 2007 Edition through 2008 Addenda [16]. The methodology and criteria to perform the flaw evaluation herein using ASME Code Section XI "2001 Edition through 2003 Addenda [6]" and "2007 Edition through 2008 Addenda [16]" are identical and the material properties are the same for both the Code Editions [References 4 and 1 7]. The Code Edition has no impact on the analytical results. Mr. Mark Sharp August 04,2017 Report No. 1700609.401.R1 Page 19 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Fatigue and SCC growth analyses has been performed for the flaw at WIB-228 because the exact root cause of the cracking is not confirmed. Vibration was not included since it was ruled out as a possible cause. Temperature data collected along the similarly affected RHR suction line for Unit 2, which shows significant thermal stratification cycling near the location of the identified flaw, has been used to perform the flaw evaluation for Unit 1. The results show that at the weld location, it would take 35 months for the initial flaw of 0.2 inch to reach the allowable flaw size. The results indicate that FCG is the predominant contributor to crack growth and while the presence of SCC is considered unlikely, it is conservatively included for the purposes of this evaluation. Thus, based on the crack growth rates presented in this report, the flaw in the 45° elbow weld of the RHR suction line at DCPP, Unit 1 is acceptable for continued operation until the 1 R21 refueling outage which will not exceed a two-year duration. REFERENCES 1. SI Calculation No. 1600539.301, Rev. 0, "Evaluation of the Flaw in the RHR Suction Weld Joint." 2. Sl Report No. 1600539.401, Rev. 0, "Flaw Evaluation ofRHR Suction Weld Joint Final Report at Diablo Canyon Power Plant, Unit 2." 3. DCPP Design Input Transmittal No. DIT-50915871-001-00, "Unit 1 RHR suction line loads and configuration," SI File No. 1700609.206. 4. ASME Boiler and Pressure Vessel Code, Section II, Part D-Material Properties, 2001 Edition, with Addenda through 2003. 5. DCPP Design Input Transmittal No. DIT-50852155-003-02, "2R19 RHR Suction Weld Joint Flaw Evaluation," SI File No. 1600539.211. 6. ASME Boiler and Pressure Vessel Code, Section XI, 2001 Edition, with Addenda through 2003. 7. pc-CRACK 4.1 CS, Version Control No. 4.1.0.0, Structural Integrity Associates, December 31, 2013. 8. SI Calculation No. 1600539.304, Rev. 0, "Crack Growth Evaluation in the RHR Suction Weld Joint." 9. ANSYS Mechanical APDL, Release 14.5 (w/ Service Pack 1 UP20120918), ANSYS, Inc., September 2012. 10. SI Calculation No. 1600539.303, Rev. 0, "Finite Element Stress Analysis of the RHR Piping." 11. Sl V&V Report No. 1700609.301, Rev. 0, "Verification ofpc-CRACK 4.2 output." 12. Section XI Task group for Piping Flaw Evaluation, ASME Code, the Journal of Pressure Vessel Technology, "Evaluation of Flaws in Austenitic Piping" Volume 108, Number 3, dated August 1986. 13. ASME Code Case N-809, "Reference Fatigue Crack Growth Rate Curves for Austenitic Stainless Steels in Pressurized Water Reactor Environments Section XI, Division 1 ," Cases the ASME Boiler and Pressure Vessel Code, June 23, 2015. Mr. Mark Sharp August 04,2017 Report No. 1700609.401.R1 Page 20 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation 14. BWRVIP-14-A: BWR Vessel and Internals Projects, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals, EPRI, Palo Alto, CA: 2008. 1016569. 15. Email from Mark Sharp (PG&E) to Aparna Alleshwaram (SI), "CONFIDENTIAL: Initial EI 50915871 -OM7.ID7 Problem Statement-Indication found using UT at weld WIB-228," dated 4/28/2017, SI File No. 1700609.201. 16. ASME Boiler and Pressure Vessel Code, Section XI, 2007 Edition, with Addenda through 2008. 17. ASME Boiler and Pressure Vessel Code, Section II, Part D -Material Properties, 2007 Edition, with Addenda through 2008. Mr. Mark Sharp August 04,2017 Report No. 1700609.401.R1 Page 21 of21 Diablo Canyon Power Plant Unit 1 Residual Heat Removal (RHR) Piping Flaw Evaluation Prepared by: Apama Alleshwararn Senior Consultant Checked by: Stan Tang Associate Verified by: Richard Bax signing on behalf of D. J. Shim D. J. Shim Associate Approved by: Robert McGill, P .E. Senior Associate 08/04/17 Date 08/04/17 Date 08/04/17 Date 08/04/17 Date }}