ML050970082
| ML050970082 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 03/25/2005 |
| From: | Pace P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML050970082 (74) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 MA 25 205 10 CFR 50.55a(a)(3)(i)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
In the Matter of ) Docket No.50-390 Tennessee Valley Authority )
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI, INSERVICE INSPECTION -
REQUEST FOR RELIEF 1-RR APPLICATION OF CODE CASE N-597-1, REQUIREMENTS FOR ANALYTICAL EVALUATION OF PIPE WALL THINNING In accordance with 10 CFR 50.55a(a)(3)(i), TVA is requesting relief from specified inservice inspection requirements in Section XI of the ASME Boiler and Pressure Vessel Code for WBN Unit 1. Enclosure 1 to this letter provides request for relief 1-RR-07 for NRC review and approval for the use of ASME Code Case N-597-1, "Requirements for Analytical Evaluation of Pipe Wall Thinning,Section XI, Division 1." The request for relief is an alternative involving the use of the Code Case that has an acceptable level of quality and safety for this application.
Pursuant to the conditions of Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 13," TVA requests review and approval for application of Code Case N-597-1 for the location described. Code Case N-597-1 is conditionally acceptable for use as described in Table 2 of Regulatory Guide 1.147, Revision 13.
Condition 1 is met as TVA's Flow Accelerated Corrosion (FAC)
Program is written to the provisions of Electric Power Research Institute (EPRI) Nuclear Safety Analysis Center (NSAC) Report 202L-R2.
c PRed cne cydepagw
U.S. Nuclear Regulatory Commission Page 2 MAR 2 5 2005 Condition 2 is met by submittal of this request for relief to use the Code Case.
Condition 3 does not apply to this application of the Code Case as this is a Class 2 component.
Condition 4 is met by scheduling this component for replacement in the Cycle 7 refueling outage in conjunction with the steam generator replacement project.
Condition 5 does not apply to this application of the Code Case as the corrosion phenomenon has been determined to probably be due to FAC.
Use of this code case was requested for this same component following the Cycle 5 refueling outage and was approved in the August 27, 2004, NRC Safety Evaluation of Request for Relief 1-RR-05 (TAC No. MC1580). The safety evaluation requires additional relief for continued operation until the end of Cycle 7. The use of this Code Case was discussed with NRC reviewers in a teleconference call on March 4, 2005. As stated in the August 27, 2004,.NRC letter, the relief request is to meet the following conditions:
The licensee must demonstrate that the structural integrity and safety of the degraded elbow will be ensured by performing a stress analysis, using the set of ultrasonic testing data recorded during the Cycle 6 refueling outage and projected to the end of Cycle 7. The stress analysis should be based on the licensing basis methodology that was approved by the staff in the resolution of Watts Bar Licensing Issue 20(a), and documented in Supplement No. 16
[actually documented in Supplement 13] of NUREG-0847, "Watts Bar Safety Evaluation Report," in accordance with the provisions of Code Case N-597-1 for piping stress analysis. contains the Cycle 6 refueling outage FAC Component Evaluation Report including the ultrasonic examination data. contains details of an evaluation relative to the Code Case for the piping elbow and includes the requirements described above. Enclosure 4 contains a copy of the Code Case preceded by the applicable conditions required by Regulatory Guide 1.147, Table 2.
U.S. Nuclear Regulatory Commission Page 3 MAR 2 5 2005 TVA requests approval of this request for application of Code Case N-597-1 by January 31, 2006, prior to the predicted wall thickness tp falling below the required minimum wall thickness t~icalculated to occur in April, 2006.
There are no regulatory commitments associated with this submittal. If you have any questions about this confirmation letter, please contact me at (423) 365-1824.
Sincerely, P. ace Manager, Site Licensing and Industry Affairs Enclosures
- 1. Request for Relief 1-RR-07
- 2. FAC Component 103BE252 Evaluation Report
- 3. Evaluation of FAC Component 103BE252 - US TOW relative to ASME Code Case N-597-1
- 4. Regulatory Guide 1.147, Table 2 and Code Case N-597-1 Enclosures Cc: (Enclosures)
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303
U.S. Nuclear Regulatory Commission Page 4 MAR 2 5 2005 PLP:RNM Enclosures cc: (Enclosures)
A. S. Bhatnagar, LP 6A-C M. F. DeRoche, ADM lQ-WBN J. C. Fornicola, LP 6A-C J. E. Hinman, ADM 1B-WBN W. M. Justice, EQB 2A-WBN G. J. Laughlin, MOB 2R-WBN F. C. Mashburn, BR 4X-C NSRB Support, LP 5M-C K. W. Singer, LP 6A-C J. E. Semelsberger, EQB 2W-WBN E. J. Vigluicci, ET 10A-K K. W. Whittenburg, SP2B-C Sequoyah Licensing Files, OPS 4C-SQN EDMS, WT 3B-K M:\SUBMIT\CODE CASE N-597-1 RELIEF REQUEST C7.doc
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 APPLICATION OF ASME CODE CASE N-597-1
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 Sununary: During implementation of WBN's Unit 1 Cycle 6 Refueling Outage Flow Accelerated Corrosion (FAC)
Program, ultrasonic examination detected wall thinning in an ASME Code Class 2 main feedwater piping elbow at the steam generator Loop 2 inlet nozzle. The FAC Program identifier for this component is grid 103BE252. Currently, the thickness is greater than the required minimum wall thickness tn. Based on the ultrasonic testing (UT) thickness measurements, predicted wear rate, and analytical analysis, it was determined that the predicted wall thickness tp will fall below the required minimum wall thickness tin 13 months after startup from the refueling outage. However, analysis demonstrates the elbow meets the alternative evaluation criteria of ASME Code Case N-597-1, Section -3600.
This condition was reported in WBN's Corrective Action Program, as Problem Evaluation Report (PER) 77658. The other three steam generator elbows to the inlet nozzles were ultrasonically examined and were found acceptable.
Use of the code case was requested for this same component following the Cycle 5 refueling outage and was approved in the August 27, 2004 NRC's Safety Evaluation of Request for Relief 1-RR-05 (TAC NO MC1580). The safety evaluation requires additional relief for continued operation until the end of Cycle 7 and the relief request is to meet the following conditions:
The licensee must demonstrate that the structural integrity and safety of the degraded elbow will be ensured by performing a stress analysis, using the set of ultrasonic testing data recorded during the Cycle 6 refueling outage and projected to the end of Cycle 7.
The stress analysis should be based on the licensing basis methodology that was approved by the staff in the resolution of Watts Bar Licensing issue 20(a), and documented in Supplement No. 16 [actually documented in Supplement 13] of NUREG-0847, "Watts Bar Safety Evaluation Report," in accordance with the provisions of Code Case N-597-1 for piping stress analysis.
El-1
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 The Cycle 6 refueling outage FAC Program ultrasonic examination of the elbow found the current minimum measured wall thickness tmeas to be 0.635 inches in Row 2 on the upstream end of the elbow. This is the same location as the Cycle 5 refueling outage minimum measured wall thickness tas of 0.639 inches. Using an estimated wall thinning rate of 0.0197 inches/year (which includes a ten percent safety factor), the predicted wall thickness tp at the Cycle 7 refueling outage is calculated to be 0.605 inches.
The allowable minimum wall thickness tmi as calculated by the equation specified in Code Case N-597-1 Paragraph -3622.1(a)(1) is 0.613 inches.
This equation is essentially the same equation for calculating the minimum wall thickness based upon the allowable hoop stresses. As noted, the predicted wall thickness tp of 0.605 inches will fall below the minimum wall thickness tmsh of 0.613 inches; but, is greater than ninety percent of the minimum wall thickness tab as allowed by the provision of the Code Case. Thus, TVA is requesting review and approval for application of the Code Case N-597-1 as allowed by the conditions of Regulatory Guide 1.147 "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 13." Application of the Code Case provides an acceptable level of quality and safety for this application.
El-2
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 Unit: 1 System: Main Feedwater System Component: FAC Grid 103BE252, 16-inch nominal pipe size (NPS) feedwater pipe 45 degree elbow at the inlet to Loop 2 Steam Generator.
Code Class: 2 Code Requirement: ASME Section XI, 1989 Edition, IWA-4300 provides a process for accessing a component for continued service after a defect has been removed. This provision stipulates that where the section thickness has been reduced below the minimum design thickness, the component shall be repaired.
As an alternative, the component may be evaluated and accepted in accordance with the design rules of either the Construction Code or Section III.
Basis For Relief: Regulatory Guide 1.147, Revision 13, conditionally accepted the use of Code Case N-597-1 subject to five conditions. Some of these conditions require prior NRC review and approval to continue to use the Code Case.
Proposed Alternative: As an alternative to the requirements of IWA-4300, TVA proposes to use the provisions of the ASME Code Case N-597-1 for analytical evaluation of FAC grid 103BE252 - US TOW, subject to the conditions incorporated into the acceptance of the Code Case in Regulatory Guide 1.147, Revision 13.
Justification For The Granting of Relief: Actual wall thickness measurements were taken by ultrasonic examination on February 27 and 28, 2005 during the Cycle 6 refueling outage. A copy of the examination data report is included in Enclosure
- 2. This is the second set of thickness measurements taken for this component. The data grid points for these measurements are established by procedure and are taken from the grid pattern El-3
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 on the component surface as depicted in the sketches labeled View 1, View 2, and View 3 of .
In addition to performing the normal FAC UT for thickness on the grid, a 100 percent UT scan at the toe of the weld (TOW scan) was performed. The lowest thickness reading and its location were recorded. The elbow is basically lying in the horizontal position. The location of the "Al" (grid origin) is approximately at the top of the horizontal portion of the elbow on the upstream end. With the grid origin located at approximately the 12:00 position on the upstream end of the elbow, the number "1' row is located on the elbow at the TOW. This actually places the column "E" at the intrados and the column "N" at the extrados. With the configuration of this portion of the pipe loop, it would be expected that wear would occur in the extrados portion of the subject elbow and would be elongated in the direction of flow.
The ultrasonic examination data was entered in the WBN FAC Manager Program and a FAC Component Evaluation Report was generated. Enclosure 2 contains the evaluation which includes the previously discussed sketches and ultrasonic examination data report. The minimum measured thickness, toss, was measured to be 0.635 inches.
This compares to the Cycle 5 refueling outage minimum measured thickness t.eas of 0.639 inches measured on September 16, 2003.
The Wear, Wear Rate, Remaining Service Life, and Status were then determined separately for each end of the elbow and for the middle portion by dividing the elbow into three regions, in order to more accurately deal with counterbore effects.
Region 1 is identified as 103BE252-US TOW which includes only the upstream Rows 1 and 2. Region 2 is identified as 103BE252-DS TOW which includes only the downstream Row 10. Region 3 is identified as 103BE252-BAL which includes Rows 3 through 9. It was decided to evaluate the Region 1 and 2 rows as a band like a pipe (because of the counterbore) and Region 3 as a blanket since that region was more like an elbow. The results for E1-4
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 Region 1 (the US TOW region) where the minimum measured reading t was obtained in both the Cycle 5 and Cycle 6 refueling outages show a wear rate of 0.0197 inches/year; Region 2 (the DS TOW region) shows a wear rate of 0.0149 inches/year; and Region 3 (the BAL region) shows a wear rate of 0.053 inches/year.
The Cycle 6 wear rate of 0.0197 inches/year compares to a wear rate of 0.025 inches/year as predicted from the Cycle 5 refueling outage data.
During the Cycle 6 refueling outage, 22 Main Feedwater elbows (one 18-inch Schedule 80 (0.938 inch Tno.) and 21 16-inch Schedule 80 (0.844 inch Tnom)) were inspected. For these 22 elbows, the average wear rate was 0.0285 inches/year; the maximum wear rate for 103BE252 was at 0.0869 inches/year (the "Loc PTP" method); and the minimum wear rate was 0.0113 inches/year.
The wear and the limiting thickness (the lowest UT reading) were used to calculate the remaining service life, "Rem Life" in years. The FAC Manager Program calculated the Rem Life to be less than the 1.5 years required to operate until Refueling Outage 7. This information is shown in .
The UT data and the physical configuration of the pipe in question indicate there was an extra long counterbore applied to the upstream end of the elbow during construction and that Band 3 is a transition area between counterbore and non-counterbore. The thickness required after counterbore is a nominal 0.758 inches with a tolerance of +0 inches and -0.005 inches. This thickness is only 0.037 inches thicker than the Tacpt value of 0.721 inches. Also, during construction the weld was prepared manually on the outside diameter surfaces for nondestructive examination (NDE). As seen from Enclosure 2, even though the location and orientation of the "thin" areas are not typical of FAC wear, WBN is experiencing FAC wear in some places of Row 2 as shown in Enclosure 2, View 2 and pages 12F and 12G of the UT data report for the locations found during the Cycle 5 refueling outage plus the new El-5
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 area shown in View 3 and page 12A of the UT data report.
Based on minimum measured wall thickness (tmeas) data obtained during the Cycle 6 refueling outage and the wear rates computed (including ten percent safety factor), the minimum projected wall loss thickness (tp) at the time of the Cycle 7 refueling outage was determined to be 0.605 inches. A structural evaluation providing justification that tp satisfies ASME Code Case N-597-1 was performed. This evaluation is included in Enclosure 3. As discussed in Enclosure 3, all pertinent elements of the Code Case N-597-1 evaluation have been applied to address the projected thickness of WBN FAC Component 103BE252-US TOW at the time of Cycle 7 refueling outage.
For each element, acceptability has been demonstrated in accordance with N-597-1 requirements. This evaluation also includes the requirements of the August 27, 2004 safety evaluation. Thus it is concluded that this piping component is acceptable for service without repair or replacement until the WBN Cycle 7 refueling outage. The elbow will be replaced in conjunction with the steam generator replacement project during the Cycle 7 refueling outage.
The following additional information is based upon suggested content in an NRC memorandum, dated August 6, 2003, for relief requests where the pipe wall thickness is less than tmz-
- 1. Markup of piping isometric showing location where piping is less than tab:
Enclosure 2 contains four sketches. The Grid Sketch is a general view of the main grid and any upstream and/or downstream grid associated with it during the inspection.
This sketch is intended to provide the inspector and the evaluator a picture of the component being considered, its location, its orientation, and its upstream/downstream components. The grid sketch does not contain grid lines.
E1-6
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 The sketches labeled View 1, View 2, and View 3 depict the thin areas/locations identified by the ultrasonic examination data. Also noted is the growth of the thinned areas/locations since the initial inspection performed during the Cycle 5 Refueling Outage.
- 2. Affected System Main Feedwater System
- 3. System normal operating temperature = 440 degrees Fahrenheit (F)
Normal operation pressure = 1145 pounds per square inch (psi)
Design Pressure = 1185 psi
- 4. Pipe size and nominal pipe wall thickness (tnom) :
16-inch Schedule 80 (tnom = 0.844 inches)
- 5. Code-allowable ten, The allowable minimum wall thickness tmin as calculated by the equation specified in Code Case N-597-1 Paragraph -3622.1(a)(1) is 0.613 inches. This equation is the same equation for calculating required minimum wall thickness for pressure as is specified in the Main Feedwater Construction Code of Record.
The pressure used in the -3622.1(a)(1) equation is the main feedwater system design pressure of 1185 psi per the definition of (P) in Code Case N-597-1.
- 6. Current thickness and date measured:
tmeas = 0.635 inches measured on February 27 and 28, 2005.
- 7. Estimated wall thinning wear rate:
0.0197 inch/year (includes ten percent safety factor).
E1-7
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1
- 8. Predicted wall thickness (tp) at Refueling Outage 7:
0.605 inches.
- 9. Discuss how pressure spikes associated with anticipated system transients are accounted for in establishing tam.
As discussed in Question 5 above, the system design pressure of 1185 psi is used to compute ti, in accordance with the equation in Section -3622.1(a)(1) of Code Case N597-1.
This pressure bounds the pressure used for the faulted condition Check Valve Slam water hammer transient.
- 10. Provide licensee's basis for determining the wear thinning rate.
The Wear, Wear Rate, Remaining Service Life, and Status were determined separately for each end of the elbow and for the middle portion by dividing the elbow into three regions, in order to more accurately deal with counterbore effects. Region 1 is identified as 103BE252-US TOW which includes only the upstream Rows 1 and 2. Region 2 is identified as 103BE252-DS TOW which includes only the downstream Row 10. Region 3 is identified as 103BE252-BAL which includes Rows 3 through 9. It was decided to evaluate the Region 1 and 2 rows as a band like a pipe (because of the counterbore) and Region 3 as a blanket since that region was more like an elbow. The results for Region 1 (the US TOW region) where the minimum measured reading tmeas was obtained in both the Cycle 5 and Cycle 6 show a wear rate of 0.0197 inches/year; Region 2 (the DS TOW region) shows a wear rate of 0.0149 inches/year; and Region 3 (the BAL region) shows a wear rate of 0.053 inches/year.
- 11. Provide licensee's criteria for repairing or replacing piping and the basis for the criteria.
E1-8
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1 TVA's FAC Program allows three options if the predicted remaining service life is less than the amount of time until the next inspection.
These options are: a) shorten the inspection interval; b) perform a more detailed stress analysis to obtain a more accurate value of the acceptable service life using the evaluation methods specified in Civil Design Standard DS-C1.2.5, "Structural Evaluation of Wall Thinning in Pipe Due to Flow Accelerated Corrosion" and, c) repair or replace the component. In addition, for piping and components with predicted wall thinning that satisfies the acceptance criteria for continued service, the program requires monitoring areas containing FAC degradation during successive examinations until repaired or replaced with a FAC resistance material and determining the frequency of future examinations by the FAC wear rate or remaining service life calculated from examination data.
For this application, this elbow will be replaced during the next refueling outage in conjunction with the steam generator replacement project. The program requires that when the predicted remaining service life is shorter than the time until the next outage, the appropriate corrective action document be initiated. WBN PER 77658, was initiated during the Cycle 6 refueling outage.
WBN's planned action regarding FAC grid 103BE252, pending approval of this request, is to replace this elbow in conjunction with the steam generator replacement in the Cycle 7 refueling outage.
- 12. Discuss what evaluation methods and criteria the licensee plans to use for performing analytical evaluations of pipe wall thinning in Class 1 carbon steel piping subjected to FAC.
WBN has no Class 1 carbon steel piping within the FAC Program.
E1-9
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR RELIEF 1-RR-07 APPLICATION OF CODE CASE N-597-1
- 13. Discuss what evaluation methods and criteria the licensee plans to use for performing analytical evaluations of pipe wall thinning in non-Code Class 1 carbon steel piping subjected to FAC.
Analytical evaluations are performed in accordance with ASME Section III, B31.1 and the Electric Power Research Institute (EPRI)
Nuclear Safety Analysis Center (NSAC) 202L Revision 2 "Recommendation for an Effective FAC Program," for non-Code Class 1 carbon steel piping subjected to FAC. A more detailed discussion of the background and approach for analytical evaluation of FAC in WBN main feedwater piping is provided in Enclosure 3.
El-10
ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 PERTINENT PAGES RELATED TO FAC COMPONENT 103BE252 EVALUATION REPORT
a Watts Bar Nuclear Plant Unit I Outage: Ij1Quage 6 FAC-Component Evaluation Report System: FW UT Analvals Innuts Uime Dia Pipe Thorn Taccpt Ta A -_ SFrlnnel Scan mnm FIL ..o Subcornponeri (yar) (in) Class n) (1n) Basis Uj SF Od M S4Cum CSTOA (n
.08a1 103BE252 1.34 16 t 0.844 0.721 N 1.000 1.10 0.752 . 0.635 l 0.709 0.535 H 103BE252-OAL T.53 16 0.844 0.721 1.00 110 0.752 7 0.752 (n 103BE252-DS TW 53 16 0.844 0.721 1. 1.10 Q i30 0.709 0.709 t1 1038E252-US TOW 18.63 16 0.84-4 0.721 II. 1.10 0.691 0.635 0.635 UT Analysis Outputs cHH T Wear Rates l Reined UtA Calculated Sample Subcomponent Med UTA Wear Calc Delinlion Wear t-es
-Corr i- WR Tlim - -n Rem k NSI Eifpslon Status tau 00M Mi
( nin fy I (Innow) r r) (in) (yr) (yln) z 1030E252 Loc PTP evious inspection 9)30R20 0.108 0.079 0.087 l 0.0889 0.835 .0.99 5 N FAIL 11038E252-BAL BLANKET Steps RadiaIIAylal 3t4 0.413 0.0484 0.0531 0 0.58 6 N FAIL (P 14(- 0 103BE252-DSTOW MaxBAND Band StartjEnd 1/1 0.115 0.0135 0.015 0.0149 0.709 .0.e0 5 N FAIL J (I]
103BE252-tUSTOW MaxBAN) BandStart/End 1U2 0.153 0.0179 0.020 0.0197 0.636 4.36 3 l N FAIL JUT Analysis Comments I H l Subcomponent E~ekuation Comments RUTA Comments 1103BE252 xMe Point to Point performed on Elbow Exduding Raldial Bands 1,2,10 due to Counterbore. RUTA Performed due to LUT Seen Value. GAD 3l/05 1038E2524JAL . THISAREA IS THE ELBOWEXCLWDINC( THE TOW ROWS #1#2AND #10-NOT I0 :*
IMPORTED TO THIS GRID GAD 3/1t05 _ _
103BE252-DS TOW THIS AREA IS THE TOW ROW a1a-EXCLUDING ALL OTHER GRID POINTS - NOT RUTA Performed on Radial Row #10 for THIS GRID_TOGAD 31105 CMIs using the Wear Rate (momn the Max Band Cale and appyIng It to the TOW Scan
_Values. - GAD 3/S05M H 1103BE262-US TOW IRUTA PERFORMED ON RADIAL ROW #1 AND #2 FOR CIVLS USING THE WEAR RUTA PERFORMED ON RADIAL ROW 1 I-i RATE FROM THE MAX BAND CALC AND APPLYING IT TO THE TOW SCAN VALUE - AND #2 FOR CIVILS USING THE WEAR GAD 3J1105 RATE FROM THE MAX BAND CALC AND APPLYING IT TO THE TOW SCAN VALUE -
A ( p '7 S ,IGAD 3/1/05 p w(SAgn ndDUSl .m/A) R ASW (sin rnf dl)
Tuesday. tmuh 0r.2005
Excluded Values In () Watts Bar Inspection Data Report Sca VWVd Subcomnponent: 1038E252 Upstream:
Inspected 2/23/2005 Outage: 0 Tmaxmeas-grld (coords): 1.285 (7.1) Downstrem: 0.709 Tnom: 0.844 Tminmeas-grkdlTnorn: 89.1% Trrnrmeas-grid (coords): 0.752 (3.C) Entire Subcomp: 0.635 Radial Coordinatos - H A 8 C 0 E F G H I .1 K L M N F 1 (0.700) (0.702) (0.737) (0.747) (0.736) (0.737) (0.768) (0.783) (0.751) (0.734) (0.765) (0.782) (0.754) (0.731) 2 (0.778) (D.785) (0.750) (0.754) (0.754) (0.751) (0.741) (0.789 (0.776) (0.762) (0.744) (0.702) (0.091) (0.721) 3 1.074 1.067 0.752 0.764 0.785 0.781 0.783 1.130 1.03B 0.897 1.075 1.108 1.114 1.095 0 4 NRT NRT NRT NRT NRT NRT NRT NRT NRT NRT 1.110 1.099 1.098 1.108 w 5 NRT NRT NRT NRT NRT NRT NRT NRT NRT NRT NRT 1.088 1.076 1.090 a NRT NRT NRT NRT NRT NRT NRT NRT NRT 1.173 1.108 1.080 1.076 1.081 7 NRT NRT NRT NRT NRT NRT NRT NRT 1.285 1.185 1.092 1.074 1.071 1.081 0 10 j 8 1.200 1.193 1.158 1.089 1.1S8 1.183 1210 1.221 1.254 1.129 1.078 1.060 1.080 1.074 9 1.051 1.008 0.995 0.988 0.935 0.993 0.978 0.949 0.890 0.843 0.889 0.899 0.941 0.935 10 (0.800) (0.806) (0.813) (0.787) (0.732) (0.777) (0.730) (0.743) (0.764) (0.745) (0.781) (0.813) (0.793) (0.790)
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IA V oto1 0
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- 0. Maneaff2001, Rote.,. 013106 P.1FAC 3II/2g~1:*~ u.l(
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Excluded Values In () Watts Bar Inspection Data Report - Sean Vabes Subcomponent 103BE252 Upstream:
Inspected: 2/2312005 Outage: 6 Tmaxmeas-grid (coords): 1.285 (7.1) Downstream: 0.709 Tnom: 0.844 Tminmeas-grldlTnom: 89.1% TmInmeas-grld (coords): 0.752 (3,C) Entire Subcomp: 0.635 Rad al CnnrdInsm -
0 p a F I (0.740) (0.741) (0.718) 2 (0.78B) (0.735) (0.771)
I 3 1.135 1.118 1.119 0 4 1.134 1.114 1.148 w 5 NRT NRT NRT a NRT NRT NRT 00 to1 7 1.128 1.125 1.182 8 1.116 1.102 1.172 1 0.968 0.085 1.003 z 10 (0.548) (0.544) (0.823) I"Hi I ~.tTi~ 0 En tJ1 C),
0 CiJ H
ti Hj -
Pahd Owe 3WONe '1-ST PM Pd01.d O~
AC 316120 1117P MManager2003, Release 013106Pa Pt g22of22
Excluded Values in (I Watts Bar Inspection Data Report SC" Vale Subcomponent: 103BE252.BAL Inspected: 2123/2005 Outage: 6 Tmaxneas-grid (coords): 1.285 (7,1)
Tnom: 0.844 Tminmeas-griddTnom: 89.1% Tminmeasgrld (coords): 0.752 (3,C)
Radial Coordinates-A B C D E F G H I J K L M N F 3 1.074 1.087 0.752 0.764 0.765 0.781 0.783 1.130 1.038 0.897 1.075 1.108 1.114 1.095 4 NRT NRT NRT NRT NRT NRT NRT NRT NRT NRT 1.110 1.099 1.098 1.108 w 5 NRT NRT NRT NRT NRT NRT NRT NRT NRT NRT NRT 1.088 1.07B 1.090 0 e NRT NRT NRT NRT NRr NRT NRT NRT NRT 1.173 1.108 1.080 1.076 1.081 w 7 NRT NRT NRT NRT NRT NRT NRT NRT 1.285 1.165 1.092 1.074 1.071 1.081 8 1.200 1.193 1.158 1.039 1.158 1.1B3 1.210 1.221 1.254 1.129 1.078 1.060 1.060 1.074 9 1.051 1.008 0.995 0.988 0.935 0.993 0.978 0.949 0.890 0.843 0.809 0.899 0.941 0.935 0I0 C!H V -a c I7I Prnitd On, :It200512:1210 PM PtI~1sd On:
311 2006 12 M62~6~
PMFAC kAW r 2003, Roiesse 013106 e. 16 Pa1012
Excluded Values In () Watts Bar Inspection Data Report Subcomponent: 1038E25248AL Inspected: 21312005 Outage: 6 Tmameas-grid (coords): 1.285 (7.1)
Tnom 0.844 TmlnmeanridlTnom: 89.1% Tmfnmeas-gdd (coords): 0.752 (3,C)
Radial Coordinates -
o P Q F 3 1.135 1.116 1.119 4 1.134 1.114 1.145 1 5 MRT NRT NRT z 0 5 NRT NRT NRT W 7 1.128 1.125 1.182 8 1.110 1.102 1.172 9 0.988 0.985 1.003 ti\)
V 10 C:H
~f-i Panted Onl: 3IV22ll12X2:33 PM Printe IAC On:
3t/21S112:12:3 Manaam PU f2003. Rdefase 0l13113Pa o PaV202
Excluded Values In ()
Watts Bar Inspection Data Report
- Sean Values Subcomponent: 103aE2524US TOW Upstream:
Inspected: 2/23/2005 Outage: 6 Tmaxmias-grd (coords): 0.789 (2.H)
Downstream:
Tnom: 0.844 Tminmeas-gridrTnom: 81.9% rmnnmeas-grld (coords): 0.691 (2.fM Entire Subcomp: 0.635 Radial GoOrrtinates H
A R C D E F G H I J K L M N F 1 0.700 0.702 0.737 0.747 0.736 0.737 0.708 0.783 0.751 0.734 0A.S 0.782 0.754 0.731 2 0.770 0.765 0.750 0.754 0.754 0.751 0.741 0.789 0.776 0.762 0.744 0.702 0.691 0.721 CH 0 N w w
0 z l V 0-H Prined On: UWN205 1:1:23 PMl Pr~ntedOn:
Unugaf 3112005 1::23 PUFAC 2003, Release 013108P6.61 PAP11`2
Excluded Values in () Wafts Bar Inspection Data Report
- Scan Vkiues Subcomponent 103BE252-IJS TOW Upstream:
Inspected: 2)2312005 Outage: 5 Tmaxmeas-grld (coords): 0.759 (2,1-) Downstream:
Tnorn: 0.844 Tnhnmeas-gridIrnom: 81.9% rmlnmeas-gid (coords): 0.691 (2,M) Entire Subcomp: 0.635 Radial Coordinates -
0 F 1 0
0.740 p
0.741 0.718 !
0.768 0.735 0.771 t2j hi 1 2 0
w z
SH' 0
v 0x ~
-J En
- z-n NJ Hi~j tzH Pewned ON: 3RrV6 t1:27 PM edCUtt2 Cii:
er~ inagwr2003. R eleassa1islesP M7PMFA Slt2O g 202 Psue2 o
Excluded Values In {) Watts Bar Inspection Data Report
- Sn ValWUs Subcomponent: 103BE252-DSTOW Upstream:
Inspected: 2/23/2005 Outage: 8 Tmaxmeas-grtd (coords): 0.848(10,0) Downstream: 0.709 Tnom: 0.844 TmInmeas-grfdlTnom: 88.5% Trmlneas-grid (coords): 0.730 (10.G) Entire Subcomp:
-~ Radial Coordinates -
A B C D E F G H I J K L M N F 10 0.800 0.808 0.813 0.787 0.732 D.777 0.730 0.743 0.764 0.745 0.761 0.313 0.793 0.790 l
0 w
t~1 z
~0 0 V U, CH ti-I N)
PTIsed On: 3w1n200 tno:11 PM P~
k~t~
0n SW 200S M1:
O;1~
Pt~9FAC ai g r 2 MO,R elweas e1tMi19 C Pap i v2
Excluded Values in () Watts Bar Inspection Data Report Cc**nVskies Subcomponentd 103BE252-DS TOW Upstrearn:~
Inspected: 2/2312005 Outage: 6 Tmaxmeas-grid (coords): 0.848 (10.0) Downstream: 0.702 Tnorw 0.844 Tmlnmeus-grfdTnom: 86.5% rmlnmeas-grd (coords): 0.730 (10,G) Entire Subcomp:
Radial Coordinates -
O Q F 10 OM 0.844 0.823 t1-0 w
001 tTI z
t:! t-4~ 0 I M ZH EnH 2 0 l Cn
~1<W~-
0 HiF1 Prheed On, 3l1V20 IM1:-1 PUA Pr~
ed anever 20"3. R eim s 013104
- W.21 M
PMFAC On:31112004 a e o421 Pe"2
tto z0tTI tal HI Ht2i ~u 0 un CZ:
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I 0
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-I-
ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 WATTS BAR NUCLEAR PLANT UNIT 1
- MONITORING PROGRAM FOR FLOW-ACCELERATED CORROSION TI-31.021 GRID ID# 103BE252 CL STEAM GEN. #2
\\ 1-TW-3-113 I a
I LOCATION SG #2 REF. DWG. E-2879-TC-2 I I
AZ 157e SYSTEM 03 Feedwater I
(+Ia E2-14
ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 TENNESEEVSALLEY DMS UL7RASON7C CALIJBRATION MEORT AMrlORMIY C.ALZRATION NO. NIA PROJEMr WBN tmNr I-ALRATIONDATE OZ-Z7-0s-PROC: I N lUT- 26 REJ T.C-.:23 CALBLOCKCNO.: o} - 74f7z
. _ ~~~TYPB: qw4 5syc rl DtE6 INSTRUhlENT/ TRNSDUCER DATA CAL BLR. MEM P.: J ,00Z THERMOMETERS/N: 562776 LNSTRUMENT MANUF.: KBA THERM. CALDUE DATE: 12-21-05 SERIALNO.: M36021 DUE DATE: 7.26-05 COUPLANT: ULTAGELI TRANSDUCER MANUF.: KBA BATCH NO. 01225 SERIALNO.: 0 // 19 SIZE: *33t ar FREQ.: 13. 0 xnbz INhUMENT SUNGS CABLE TYPE FDCMD LENGTH 48" PROBE: f=2 _
_TlUCKCCAL: IPTUl 2Pno RANGE 7.od a inches
______ __VELOCI~rY. o23 gr in~a
= =_ == = = = A GAIN MMz D M TCG MODE: SINGLE
=========_ P DUAL t9 -
========== L RECTIFY: POS}IVE O
_=__====== I NGATWVE 3 U AMPLITUDE: NORMAL D SCALE 0
_ ________=__=______=____E__
DISPLAY WIDTHi 'Z-OO Inches CALIBRATIONThES INITIAL CAL TNME IA-d (I) VERIPICATIONTMIES REF.REFLECTOR BIE GAIN 7'Y dB 1) 7/W 2) 17 u13) 1 7Z4)1 AMPL /00 % METALPATHc o inches 5) A": 6) -. 1'7) -4 8) %
FINAL CALTIME / s (2) CONIPNJGRID(S) XAMNPIED TEMP REF. REFLECTOR BIB GAIN 74 dB I c3 R E 2r z 0 7 or AmIPL Ijo 91 MErALPATIL /,j3Yn-cbes D p.oj--j Aviy 87_ f COMMENTS: U1C6 T.J.-3 1.21 fRSpjgqws 7&XJ-hA 6)~D _________
ANIL' EXAININ' _ LVL DATE:
REVIEWER. .. (.. 4 LvL. PAGE Of 1 E2-15
ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 1
0112?12005 ASSIGNMENT SHEET WALL THINNING MONITORING PROGRAM FOR FAC.
CYCLE06 CAVITATION, MICROBIOLOGICALLY INDUCED CORROSION AND GENERALIZED CORROSION I
GRID Number: 1 03BE252 SYSTEM#: FW TVA Class: REFDWG: E-2879 IC-2 NDE Procedure: N-UT-26 Flow Diagram:
Grid Extents: Scan Method Point to Point Inspection Method: F ComponentType: ELBOW Material: PI Previous UT: Yes (See Attached)
Diameter. 16 Previous UT T Min: - 0.639 Schedule: 80 LOCATION: SC#2 AZ 157 T-Map: - 0.721 Exam WLJ'AL nPf - f Z IV fK-1 CCOWA) f-,:
Remarks: khFkF- 14e'6 . TmAPtp RD-j It' 1 -J Component Remarks EXAMINER , EXAMINER 2:
RWP NUMBER:
DOSE: TIME REQUIRED EXAMINER COMMENTS:
E2-16
ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 103E52S2.dat
File Header------------------------------
P c. rile Name 103BEo2 .dat Gauge File Nane s 103BE252 Description I Memo Coument Creation Date 02/2712005 Date Last Saved 01/01/1997 Probo Cal. Stnt. I Temp. Comment Inspector company Instrucent Type :M52 Xnstrument S.N. COYCRD
~in. Alarm VAL. 0.179 Max. Alarm Val. 0.000
% Loss Alarm Val. I 0.00 % Growth Alarm Val. 0.00 Abs. Loss Alarm Val. I 0.000 Abs. Crowth Alr-s Val. 0.000 units t INCH Velocity (in/us) : 0.2322
--ile-------------------------FPlo Statistics--------------------------------
NUmber of Readings 126 Nunber of Empties 0 Number of Obstructs 44 Number of Attachmests : 0 Range a 0.594 Mean 0.937 Medfaan 0.988 Standard Deviation a 0.174 Hinirun Value 0.691 Hinimrum Value Loc. 2 a x Haximum Value 1.285 Maximum Value Loc. 7: x Minimau Value Alarr= a 0 Maximum Value Alarms 0 Percent Loss Alarms 0 Percent Growth Alarnms 0 Absolute Loss Alarxa I 0 Absolute Gtowth Alarms i 0
% and Abs. Loss Alarms 0 O % and Abs. Grow h Alarms: 0
--------------------- File Coments----------------------------------
A P. CASEY I WBNlC6 ja C , H-vT-26 R23 K D THM 562776 12215 L E a UTOL1I 01225 r F TI 31.21 N 0: 0:
H: P:
Headar Page 1 E2-17
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00 0.74.7 0.730G 0.737; 0.706 0.763 0.71 0734 5 0.72 074 0.731j 0.740 0.77 0.765 C 0.7U7 0.754 0.75ti 0.74t 0.78 0.776i 0.782 0.744 0.702 0.M9 0.7211 1074 1t7j1 0.752j 0.764 0.76S 0.781W 073 1.13 1f6j 0a97t 1.D75 1.10. t..4. 1.tO'i 1.135 ObstrObd O e O1f O*
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_ D1 .1D061 j 086 0.5i 0.9 076 04 0.89 843 0.869 0-899 0.S41i 0.951i < H 0.8v0 0.C6 0.8131 0.787 0.732 0.777 0730 0.743 0.76i4 0.745i 0.761 0.813 0.793 0.790 0.46 M pd t-3 iA C4 tiY t CI M zH Etid3 M t r,
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TVA PROJECT: t A -SYSTEM 03 FZZitN11r I Orfice of NucleAr Power Unit: I I cZctCt6 o____
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TVA PROJECT: W8N SYSTEM- 03 ME3wArER REPORT NO.:
Offlce of Nuclear Power Unit: I c>ie ¶V6 NO.: JO3REZ5
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TVA PROJECT: W8N -SYSTEM; 03 -Geta l170 REPORT NO.-
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TVA PROJECT: WM' SYSTEM: 0-3 FM3W47C72. REPORT NO.:
Office of Nuclear Power lUnit: Icl NO.: 903E2SZ
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BY:_ _ LEVEL: -.- Z DATE: 02-Z8--- PAGE 126( OF4 TVA 19669 (ONP-"SI
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 Background and Objective:
This evaluation addresses flow accelerated corrosion (FAC) induced wall thinning for WBN FAC Component 103BE252, an American Society of Mechanical Engineer (ASME)Section III, Class 2, 16-inch Schedule 80 (i.e., tnom = 0.844 inches), elbow located in Loop 2 of the Unit 1 main feedwater system. This elbow is fabricated from SA-333 GR 6 material. Design Pressure and Temperature are 1185 pounds per square inch (psi) and 600 degrees Fahrenheit (F), respectively.
Based on wall thickness data taken during the Cycle 6 Refueling Outage in February 2005 (i.e., 0.635 inches minimum measured remaining thickness) and wear rates computed using WBN FAC Program methods (including ten percent safety factor), the minimum projected wall thickness (tp) at the time of Refueling Outage 7 (i.e., 1.5 years after Refueling Outage 6) was determined to be 0.605 inches. This evaluation provides technical justification that tp satisfies ASME Code Case N-597-1 requirements. Each pertinent section of the code case will be addressed separately.
Section -3500 - Wall Thickness Acceptance Standards:
Per Section -3500(a)(1), the acceptance standard that tp (0.605 inches) must be greater than 0.875 tnom = 0.739 inches is not satisfied for this component. However, tp is significantly greater than the limiting thickness requirement specified in Section -3500(d) for ASME Class 2 components (i.e., tp = 0.605 inches > 0.2 tlom = 0.169 inches). Thus, further evaluation is performed in accordance with Section -3600, "Analytical Evaluation for Class 2 and Class 3 Piping Items," as permitted by Section -3223, "Acceptance by Engineering Evaluation." Each pertinent portion of Section -3600 is addressed below.
Section -3600 - Analytical Evaluation for Class 2 and Class 3 Piping Items Section -3621 - Evaluation of Pipe, Elbows, Branch Connections and Reducers:
Per Section -3621(d), the ratio (R)/tp = 8.0 inches / 0.605 inches = 13.2 is less than 50. Thus, the potential for buckling requires no further evaluation.
E3-1
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 Section -3622 - Thickness Evaluation:
In determining acceptance in accordance with Section -3622, the wall thickness data taken during Refueling Outage 6 is used with the wear rate (plus a ten percent safety factor) applied for the length of service expected (i.e., 1.5 years until Refueling Outage 7) to determine projected thickness. For WBN FAC component 103BE252-US TOW, a minimum projected wall thickness value (tp) of 0.605 inches has been computed for Refueling Outage
- 7. Considering the design pressure (P = 1185 psi) and allowable stress for the elbow material (S = 15000 psi @ 600 degrees F),
the equation from the construction code is used to determine the required thickness (tin) as follows:
tr 2 PDo 1185psi(16.00in) -0613n n (S +y P)- 2 [l5000psi + 0.4(1185psi)]
For acceptability, Section -3622.1(a) specifies that tp may not be less than 90 percent of tin-tp 0.605in=0987,>.900 trrn 0.613in Since the 90 percent tin criteria is satisfied, the requirements of Section -3622 are satisfied without further evaluation.
Section -3623 - Piping Stress Evaluation:
Because of unusual design loading requirements applicable to the WBN main feedwater system, the Section -3623 piping stress evaluation for WBN FAC Component 103BE252-US TOW required a non-typical approach to demonstrate compliance with the analysis of record. To better understand the approach applied to FAC Component 103BE252-US TOW, the following background discussion is provided:
Background on MFW Pipe Design Basis Structural Qualification During the plant's design phase, main feedwater piping at WBN was subjected to structural analysis for applicable loading conditions in accordance with Design Criteria WB-DC-40-31.7, "Analysis of Category I and I(L) Piping Systems." Initially, load sources included in the analysis were pressure, deadweight, E3 -2
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 bellows pre-load, thermal expansion, earthquake, and loss-of-coolant accident (LOCA) effects in various specified load combinations. TVA's standard piping analysis software, TPIPE, was utilized to predict the linear elastic response of the piping and perform stress analysis in accordance with ASME Section III, Subsection NC-3600 (Class 2) rules. The initial analysis demonstrated that the piping system design met all relevant criteria for the load conditions noted above.
Subsequent to the design, analysis and installation of the WBN main feedwater piping, it was determined that the main feedwater system, within the containment isolation boundary from the check valves to the steam generators including the piping components and supports, required evaluation for pressure boundary integrity to withstand the postulated water hammer event due to feedwater check valve slam (CVS) following a pipe rupture in the feedwater piping upstream of the containment isolation check valve. The CVS fluid transient occurs as a result of a postulated instantaneous non-mechanistic rupture of the main feedwater header upstream of the containment isolation/class break valve in the non-safety related part of the system during power operation.
Flow out of the rupture results in an initial rapid blow down of the steam generators via backflow through main feedwater piping.
Steam generator blow down is quickly terminated by closure of the main feedwater containment isolation check valve, but the rapid check valve closure (i.e., slam) produces a severe water hammer type fluid transient. This event is considered a faulted plant condition but is assumed to occur concurrently with a safe shutdown earthquake (SSE). Thus, an additional analysis was required to demonstrate pressure boundary integrity of the main feedwater pipe inboard of the isolation check valve for CVS plus SSE loads to ensure safe shutdown and containment integrity.
Because of the severity of the CVS plus SSE event, linear elastic system analysis and associated acceptance criteria resulted in exceedance of faulted condition allowables for the main feedwater piping and supports. Because there were no practical modification schemes that would sufficiently increase the structural capacity of main feedwater system within the areas of concern, WBN pursued an alternative approach for qualification of main feedwater piping system based upon a nonlinear inelastic system analysis that could take advantage of the short duration of the CVS event and the overall structural capacity of the piping and support system beyond what is typically considered in an elastic analysis.
E3-3
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 The foundation for this inelastic approach is a non-linear inelastic structural analysis of three of the four main feedwater piping loops inside primary containment using the ANSYS structural analysis computer program (note that the fourth loop was qualified without analysis because it is geometrically similar to one of the analyzed loops). Fundamentally, the models developed for the ANSYS analysis are point-to-point beam models of the system that are similar in most respects to the models used in the TPIPE program. The primary difference between the ANSYS analysis software and TPIPE analysis software is the capability to consider elastic-plastic material behavior in determining general system response and the capability to simulate non-linear boundary conditions (e.g., allowing a pipe support to become inoperative when it reaches an overload condition). The ANSYS software provides the capability to consider non-linear material behavior via plastic piping elements (i.e., both straight and curved elements are available) which simulate beam behavior in the material's inelastic regime. ANSYS plastic pipe elements are geometrically defined by 2 nodes (i.e.,
the end points), pipe diameter, wall thickness and radius of curvature (i.e., if an elbow). These are not detailed three dimensional plate and shell models of the pipe or fitting.
The CVS transient loads were generated using the thermal hydraulic conditions (including pressures) in the feedwater system (modeled from steam generators to 32-inch header). The system initial conditions and CVS event parameters were modeled using the RELAP/REFORCE computer codes to generate the system transient response and produce the force time histories to be used as input in the ANSYS analysis. In addition to these CVS force-time histories, loads from SSE, Pressure, Thermal Expansion, Thermal Anchor Motion, and Deadweight were included in the ANSYS analysis and the results evaluated against Appendix F in Section III of the ASME Code stress criteria.
The stress acceptance criteria used for CVS plus SSE analysis is Appendix F in Section III of the ASME Code. Appendix F was developed to supplement the evaluation of ASME Section III, Class 1 components under faulted condition loading. Appendix F is based on the same maximum shear stress theory as is utilized for elastic analysis in Subsection NB of Section III. For the SA-333 GR 6 carbon steel material from which the main feedwater piping system is fabricated, the general membrane stress intensity limit (Sm) was determined to be 0.7 of ultimate stress (Su) per Appendix F from the 1980 Edition of ASME Section III, E3-4
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 with Addenda through Winter 1982. Based on the Appendix F 1980 Edition, with Winter 1982 Addenda, the local membrane stress intensity limit would be 1.5 Sm; however, TVA recognized that in Appendix F, 1983 Edition of ASME Section III, the local membrane stress intensity limit for plastic analysis specified in F-1341.2 (Plastic Analysis) was changed to 0.9 Su. This limit is significantly less than the 1.5 Sm limit on local membrane stress intensity allowed in Appendix F, 1980 Edition of ASME Section with Addenda through Winter 1982 (i.e., 1.5 Sm = 1.05 Su for SA-333 GR 6). Therefore, a local membrane stress intensity limit of 0.9 Su was conservatively applied for the WBN CVS plus SSE analysis to preclude primary stresses from exceeding the specified ultimate capacity for the material.
Using this approach, the WBN main feedwater piping was structurally qualified for the CVS plus SSE faulted load combination using a nominal thickness as appropriate for 16 and 18-inch Schedule 80 pipe.
For the CVS plus SSE analysis performed, the methodology and stress allowable were reviewed by the NRC under Watts Bar Licensing Issue 20(a) and approved in Supplement No. 13 of NUREG-0847.
Thus, there are two companion analyses for design basis structural qualification of WBN main feedwater.
- Linear Elastic Analysis for All Load Combinations Without Check Valve Slam Using TVA's*TPIPE Program and Standard ASME Section III, Class 2 Stress Allowables
- Nonlinear Inelastic Analysis for the Faulted Condition Load Combination Including Check Valve Slam Using the ANSYS Program and ASME Section III, Appendix F Stress Limits Accordingly, for structural evaluation of FAC degradation in WBN main feedwater, including FAC Component 103BE252-US TOW, two separate methodologies are required in order to demonstrate Construction Code of Record compliance as required by Section -
3623.
E3-5
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 A. Linear Elastic Analysis Loads Without Check Valve Slam For the linear elastic evaluation of the standard load combinations that were utilized for structural analysis using the TPIPE computer program, the process is described as follows:
- 1. Acceptance value is established for the component, based upon the remaining stress margin to Code allowable stress acceptance criteria taken from the TPIPE analysis of record.
- 2. Using current wall thickness inspection results for the piping component, a controlling wear rate (including safety factor) is computed in accordance with established methods. The remaining thickness for each data point is projected at least through the next inspection period (generally the next refueling outage).
- a. If the minimum projected thickness is greater than acceptance value for that component, the component is considered acceptable. No further evaluation is required until the next required inspection.
- b. If the minimum projected thickness is less than acceptance value for that component, the wall thickness data and wear rate (including safety factor) are subjected to further structural evaluation.
- 3. If further structural evaluation is required, reduced section properties (i.e., average thickness, remaining metal area, pressure area, principal axis moments of inertia, neutral axis shift and principal axis section modulus) are calculated using distributed minimum thickness results for the next inspection cycle that are based on projection of current thickness inspection data using the wear rate (including safety factor) determined in Step 2.
- 4. The stresses for each applicable load combination are recomputed using the system design pressure and resultant moments extracted from the TPIPE analysis of record in conjunction with the reduced section E3-6
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 properties, revised stress intensification factors (based on average cross section thickness where appropriate), and pertinent piping analysis stress equations for ASME Section III, Class 2.
- 5. Recomputed stresses are then compared to Code allowable stresses, in accordance with the WBN Design Criteria WB-DC-40-31.7, to verify acceptability for the standard load combinations.
B. Nonlinear Inelastic Analysis With Check Valve Slam As discussed previously, the main feedwater analysis of record for the CVS plus SSE faulted load combination utilized a nonlinear inelastic technique and acceptance criteria. Because of this non-typical analysis technique and acceptance criteria, it was recognized that a modified approach would be required for structural evaluation of the CVS plus SSE load combination (i.e., modified relative to the standard approach applied for load combinations evaluated based on linear elastic analysis). Furthermore, the approach must satisfy the fundamental requirements for faulted load evaluation in accordance with Appendix F of ASME Section III, 1980 Edition with Winter 1982 Addenda.
In this regard, the possibility of rerunning the ANSYS analysis with reduced thicknesses input for the pipe beam elements was considered but was determined to be inappropriate because it would introduce an excessive degree of conservatism into the evaluation. It should be noted that the limiting structural locations in the MFW system inside containment are the pipe elbows. FAC inspection results show that thinning generally involves narrow circumferential bands near the girth butt weld joints between the elbows and adjacent piping components. Because FAC inspection/evaluation results have shown that the bulk of a typical elbow remains at or above nominal thickness even after wear is considered, a simple assumption of constant thickness for an elbow element in the ANSYS pipe model could lead to overly conservative results.
An approach utilizing an ANSYS three-dimensional plate and shell model was also considered but it was determined that this technique would not be practical for application to E3-7
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 multiple locations in a short duration refueling outage inspection period. It should be noted that the WBN Unit 1 Cycle 6 main feedwater FAC inspection program included 22 elbows out of a total of 28 elbows in the four loops of main feedwater and many of these could have required structural evaluation during the outage period if the combination of counter bore geometry at girth butt welds and projected wear had indicated a violation of the 87.5 percent of nominal thickness limit allowed by the construction code.
To develop an approach that would be both technically appropriate and efficient, the theoretical basis for the ANSYS plastic elbow element that is utilized in the WBN CVS plus SSE faulted load combination analysis of record was reviewed. From this review, it was determined that the ANSYS plastic elbow element, referred to as STIF60 in the program version used for the WBN analysis of record, utilizes the elbow stress index method described in ASME Section III, Subsection NB-3685 as the fundamental basis for stress computation. This technique allows for computation of normal and shear stresses at any point around the circumference of the elbow that is subjected to two orthogonal bending moments, a torsional moment and pressure.
The ANSYS STIF60 element (referred to as the PIPE60 element in later ANSYS versions) computes stress at 8 equally spaced (i.e., 45 degrees) points around the elbow circumference for the loads resolved at each elbow end node and then applies adjustment to the stress to account for inelastic behavior.
Because the elbow stress index method is amenable to rapid numerical analysis, development of a FAC evaluation approach using this method as the fundamental basis was pursued.
First, bending and torsional moments were extracted from the CVS plus SSE ANSYS analysis of record for the elbow locations. These moments, pressure, nominal component dimensions and properties were then input to a separate analysis routine utilizing Mathcad software to generate a purely elastic but efficient solution of the NB-3685 stress index method at circumferential points coinciding with the ANSYS results. Normal and shear stress from the elastic NB-3685 solution were then converted in the Mathcad routine to principal stresses and stress intensities in accordance with ASME Section III, Class 1 stress analysis directions. It was observed that the elastic solution using this approach produces comparable and slightly conservative principal E3-8
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 stresses and stress intensities when compared to the ANSYS inelastic stress results. Thus, it was concluded that an elastic solution of the stress index method is conservative from the standpoint of stress magnitude.
To be considered as a wall thinning evaluation tool, it was recognized that the elbow stress index method must also be able to account for the reduction in cross section.
Accordingly, it may be observed that the elbow stress index method as described in NB-3685 requires two pertinent parameters as input: wall thickness and section modulus.
Wall thickness is primarily applied to compute longitudinal and hoop direction stress indices for in-plane and out-of-plane bending moments and pressure stress indices in the hoop direction. Elbow stress indices increase as thickness is reduced. Because of the fact that the elbow stress index method predicts behavior based on constant thickness over the entire component and the fact that wall thinning in main feedwater elbows is localized and confined to circumferential bands, use of an average thickness at given low thickness cross section is conservative for evaluating elbow behavior and stress in that circumferential band.
Section modulus is used in the NB-3685 stress index method to compute nominal stresses due to moment loads. Thus, use of the minimum section modulus for a given low thickness cross section is conservative for evaluating elbow behavior and stress in that location. Output from the distributed thickness section property routine that was developed for standard elastic analysis provides both an average thickness and minimum principal axis section modulus. Thus, by incorporating the ASME Section III, Subsection NB-3685 stress index method with the pertinent section properties from the distributed thickness property routine to compute stress intensity in accordance with ASME Section III, Class 1 and Appendix F stress criteria, a relatively straight forward but conservative approach was established for computing stress in locally thinned main feedwater elbows for the CVS plus SSE faulted load combination.
Because the observed thinning in WBN main feedwater elbows is localized in narrow circumferential bands near the weld joints with the bulk of the elbow remaining at or above nominal thickness values, the piping loads (i.e., moments) from the ANSYS inelastic treatment of the CVS plus SSE faulted condition remain valid for evaluation of localized thinning.
E3-9
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 The code of record ANSYS analysis of the feedwater piping was performed using the inelastic system and inelastic component analysis techniques as described in ASME Appendix F from the 1980 Edition of Section III, with Addenda through Winter 1982, Table F-1322.2-1 and is an approved methodology as stated above. It has been shown, as described above, that the Mathcad NB-3685 stress index method will produce stress magnitudes equal to or greater than the magnitude obtained when the method is applied in an inelastic fashion as in the ANSYS STIF60 element used in the analysis of record for CVS plus SSE. As such, the elastically calculated stress could be conservatively compared to the inelastic stress limit specified in Appendix F.
Nevertheless, as applied for evaluation of the WBN Unit 1 Cycle 6 degraded wall data, this approach would be considered an inelastic system analysis and an elastic component analysis as described in ASME Appendix F from the 1980 Edition of Section III, with Addenda through Winter 1982, Table F-1322.2-1. However, review of ASME Appendix F, Table F-1322.2-1 indicates that the general membrane stress limit for both types of analysis is actually the same.
Additionally, as stated earlier, the ASME Appendix F local membrane stress limit from the 1983 Edition of Section III was actually used in the analysis of record due to the fact that it was more conservative than ASME Appendix F from the 1980 Edition of Section III (i.e., the applicable Code defined in the Final Safety Analysis Report [FSAR]). In reviewing Appendix F from the 1983 Edition of ASME Section III for inelastic system and elastic component analysis, the defined stress limit value is exactly the same as Appendix F from the 1980 Edition of ASME Section III, with Winter 1982 Addenda. The allowable from both of these codes for an inelastic system and elastic component analysis produces an allowable stress limit for local primary membrane in excess of the ultimate capacity of the material (i.e., 1.5 Sm =
1.05 Su for SA-333 GR 6). Therefore, for WBN Unit 1 Cycle 6 main feedwater FAC evaluations under the CVS plus SSE faulted load combination, the same conservative 59 ksi (0.9 x Su), as approved by Supplement No. 13 of NUREG-0847, was utilized.
Thus, the general process for evaluation of FAC in WBN main feedwater elbows for the CVS plus SSE faulted load E3-10
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 combination during the Unit 1 Cycle 6 refueling outage is described as follows:
- 1. An initial acceptance value is established for the component, based upon the remaining stress margin to Code allowable stress acceptance criteria taken from the ANSYS analysis of record described above.
- 2. Using current wall thickness inspection results for the piping component, a controlling wear rate (including safety factor) is computed in accordance with established methods. The remaining thickness for each data point is projected at least through the next inspection period (generally the next refueling outage).
- a. If the minimum projected thickness is greater than acceptance value for that component, the component is considered acceptable. No further evaluation is required until the next required inspection.
- b. If the minimum projected thickness is less than acceptance value for that component, the wall thickness data and wear rate (including safety factor) are subjected to further structural evaluation.
- 3. If further structural evaluation is required, reduced section properties (i.e., average thickness, remaining metal area, pressure area, principal axis moments of inertia, neutral axis shift and principal axis section modulus) are calculated using distributed minimum thickness results for the next inspection cycle that are based on projection of current thickness inspection data using the wear rate (including safety factor) determined in Step 2.
- 4. Using bending moments, torsional moment and pressure extracted from the ANSYS analysis of record for the CVS transient with the induced bending due to pressure and neutral axis offset applied in the most conservative manner, a revised stress analysis using the ASME Section III, Subsection NB-3685 Stress Index Method (same fundamental approach as used by ANSYS) is performed based on reduced section properties (i.e.,
E3-11
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 average cross section thickness and minimum section modulus as appropriate for the location in question) to produce a set of normal and shear stresses from which principal stresses and stress intensities are computed.
- 5. Recomputed stress intensities are then compared to local membrane faulted condition allowable stresses from ASME Section III Appendix F, as appropriate, to verify acceptability.
Thus, by subjecting a main feedwater FAC component to both of the analysis processes described above, both the Construction Code of Record and the requirements of Section
-3623 are addressed.
C. Pipe Stress Evaluation Summary for Main Feedwater FAC Component 103BE252-US TOW The effect of the pipe longitudinal pressure and bending stresses was evaluated for FAC Component 103BE252-US TOW using the detailed stress analyses described above that were developed to demonstrate compliance with both the Construction Code of Record and the requirements of Section
-3623. These detailed stress analyses used as input the RFO-6 measured thickness data for the cross section in question including the wear rate (including safety factor) applied for the length of service expected (i.e., 1.5 years until Refueling Outage 7). The stress analysis methodology accounted for reduced section properties and induced bending due to neutral axis offset considering the projected thicknesses at the time of Refueling Outage 7. Also, the stress evaluation accounted for the impact on the elbow stress intensification factor (Section -3623.4) and stress indices per the CVS plus SSE analysis of record due to the projected wall thickness at Refueling Outage 7. Pipe bending and torsional moment loads were taken from the analyses of record which involve rigorous static and dynamic qualification of the piping to satisfy Construction Code of Record requirements. Although the analyses of record assume nominal thickness for all piping components, the 1.143 stress allowable increase factor permitted in Section -
3623.1 was not used in the structural evaluation. This stress evaluation showed that all applicable Code of Record ASME Section III, Class 2 equations and the ASME Section E3-12
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 III, Appendix F faulted condition (CVS plus SSE load combination) remain within required allowable stresses when considering the reduced thickness profile of the affected cross section as projected to Refueling Outage 7.
Specifically the maximum primary and sustained primary plus secondary stress ratios (applied / allowable) were found to be 0.967 and 0.834, respectively for the cross section coincident with the local region where projected minimum thickness at Refueling Outage 7 is 0.605 inches. Thus, the requirements of Section -3623 are satisfied.
Section -3624 - Evaluation of Branch Connections:
The location of FAC Component 103BE252-US TOW is a piping elbow and not a branch connection, therefore, the requirements of Section -3624 are not applicable.
Section -3625 - Evaluation for Cyclic Operation:
This component was evaluated for cyclic operation using Section -
3625(b). For the WBN main feedwater System, review of the stress range reduction factors of Table 3625-1 indicates that a factor of 1.0 would be applicable since the applicable number of full temperature thermal cycles for one operation cycle is much less than the 650. In addition, projected thicknesses at Refueling Outage 7 have been included in the development of the stress intensification factor for the Section -3623 analysis discussed above which demonstrated satisfaction of pertinent stress equations and hence suitability for cyclic service in accordance with ASME Class 2 piping requirements. For Section -3625(c),
evaluation of ASME Section III, NC-3600, Equation 11 (i.e.,
sustained primary plus thermal expansion stress) indicates that the revised stress ratio (i.e., applied / allowable) remains less than 1.0 for the analysis performed in the Section -3623 evaluation. This ensures that the local overstrain in the thinned region for the combination of maximum sustained plus thermal expansion stresses have been adequately considered.
Thus, the requirements of Section -3625 are satisfied.
==
Conclusion:==
As discussed above, the pertinent elements of the Code Case N-597-1 evaluation have been applied to address the projected thickness of WBN FAC Component 103BE252- US TOW at the time of Refueling Outage 7. For each element, acceptability has been E3-13
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 EVALUATION OF WBN FAC PROGRAM COMPONENT 103BE252-US TOW RELATIVE TO ASME CODE CASE N-597-1 demonstrated in accordance with N-597-1 requirements. Thus, it is concluded that this piping component is acceptable for service without repair or replacement until the WBN Unit 1 Cycle 7 Refueling Outage.
E3-14
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 REGULATORY GUIDE 1.147 TABLE 2 AND COPY OF ASME CODE CASE N-597-1
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 REGULATORY GUIDE 1.147, TABLE 2 CODE TABLE 2, CONDITIONALLY ACCEPTABLE SUPPLEMENT/
CASE SECTION XI CODE CASES EDITION NUMBER CONDITION.
N-586 Alternative Additional Examination Requirements for Class 1,2, 5/98E and 3 Piping, Components, and Supports, Section Xl, Division 1 The engineering evaluations addressed under Item (a) and the additional examinations addressed under Item (b) shall be performed during this outage. It the additional examinations performed under Item (b) reveal indications exceeding the applicable acceptance criteria of Section XI, the engineering evaluations and the examinations shall be further extended to included additional evaluations and examinations at this outage.
N-593 Alternative Examination Requirements for Steam Generator 11/98E Nozzle to Vessel Welds, Section Xl, Division 1 Essentially 100 percent (not less than 90 percent) of the '.
examination volume A-B-C-D-E-F-G-H must be Inspected.
N-597-1 Requirements for Analytical Evaluation of Pipe Wall Thinning, 2/01 E Section Xl, Division I (1) Code Case must be supplemented by the provisions of EPRI Nuclear Safety Analysis Center Report 202L-R2, April 1999,
'Recommendations for an Effective Flow Accelerated Corrosion Program," for developing the Inspection requirements, the*
method of predicting the rate of wall thickness loss; and the value of the predicted remaining wall thickness. As used in NSAC-202L-R2, the terms 'should' and 'shall" have the same expectation of being completed.
(2) Components affected by flow-accelerated corrosion to which this Code Case are applied must be repaired or replaced in accordance with the construction code of record and Owner's requirements or a later NRC approved edition of Section III of the ASME Code prior to the value of t, reaching the allowable minimum wall thickness, t,,-, as specified In -3622.1 (a)(1) of this Code Case. Alternatively, use of the Code Case is subject to NRC review and approval.
(3) For Class 1 piping not meeting the criteria of -3221. the use of evaluation methods and criteria is subject to NRC review and approval. I (4) For those components that do not require immediate repair or replacement, the rate of wall thickness loss is to be used to determine a suitable inspection frequency so that repair or replacement occurs prior to reaching allowable minimum wall thickness, tih.
E4-1
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 FIRST 10-YEAR INTERVAL REQUEST FOR RELIEF 1-RR-07 REGULATORY GUIDE 1.147, TABLE 2 l CODE TABLE 2, CONDITIONALLY ACCEPTABLE UPPLEMENT/
CASE SECTION XI CODE CASES EDITION
-NUMBER CONDITION N-597-1 Requirements for Analytical Evaluation of Pipe Wall Thinning, 2/01 E (continued) Section Xl, Division I (5) For corrosion phenomenon other than flow accelerated corrosion, use of the Code Case is subject to NRC review and approval. Inspection plans and waill thinning rates may be difficult to justify for certain degradation mechanisms such as MIC and pitting.
N-599 Alternatives to Qualification of Nondestructive Examination 2/98E Personnel for Inservice Inspection of Metal (Class MC) and Concrete (Class CC) Containments, Section Xl, Division 1 This Code Case may not be used when a licensee updates to the 1992 or later Edition of Section XI that requires the use of ANSI/ASNT CP-1 89, "Standard for Qualification and Certification of Nondestructive Testing Personnel."
N-606- 1 Similar and Dissimilar Metal Welding Using Ambient 6198E Temperature Machine GTAW Temper Bead Technique for BWR CRD Housing/Stub Tube Repairs, Section Xl, Division 1 Prior to welding, an examination or verification must be performed to ensure proper preparation of the base metal. and that the surface is properly contoured so that an acceptable weld can be produced. The surfaces to be welded, and surfaces adjacent to the weld, are to be free from contaminants, such as, rust, moisture, grease, and other foreign material or any other condition that would prevent proper welding and adversely affect the quality or strength of the weld. This verification is to be required in the welding procedures.
N-616 Alternative Requirements for VT-2 Vsual Examination of 6/98E Classes 1, 2, 3 Insulated Pressure Retaining Bolted Connections, Section Xl, Division 1 (1) Insulation must be removed for VT-2 examination during the system pressure test-for any 17-4 PH stainless steel of 410 stainless steel stud or bolt aged at a temperature below 110001F or with hardness above Ra 30.
(2) For A-286 stainless steel studs or bolts, the preload must be verified to be below 100 Ksi or the thermal insulation must be removed and the Joint visually examined.
(3) for nuts conforming to SA-194; removal of the insulation for visual inspection is not necessary.
(4) Prior to conducting the VT-2 examination, the provisions of IWA-5213, 'est Condition Holding Tim-es," 19B9 Edition, are to be followed.
E4-2
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE N-597-1 CASES OF ASME BOILER AND rREssutE VESSEL CODE Approval Date: September 7, 2001 See Numeric Index for expfration and any reaffirmation dates.
Case N-597-1 value, tp ,;,, may be used in determining acceptability Requirements for Analytical Evaluation of Pipe for continued service. Methods of predicting the rate Wall Thinning of wall thickness loss and the value of t, shall be the Section X3, Division 1 responsibility of the Owner.
Inquiry: What requirements may be used for analyti- .3220 Acceptance cal evaluation of Classes 1, 2, and 3 piping items subjected to internal or external wall thinning? .3221 Acceptance By Examination Piping items whose examination and evaluation re-Reply: It is the opinion of the Committee that the suits reveal that t, meets the acceptance standards of following rules may be used. -3500 or the Construction Code are acceptable for continued service. When these criteria are not met, the alternatives of -3222, -3223, and -3224 may be used.
Fig. -3220-1 shows a flow chart of the acceptance alternatives.
- 1000 SCOPE This Subsection provides requirements for analytical -3222 Acceptance by Repair/Replacement evaluation of Classes 1, 2, and 3 piping items (e.g., Activity piping and fittings) with internal or external wall thin-Piping items whose thickness is less than that required ning. These requirements are applicable to nonplanar by -3500. -3223, -3224 shall be corrected by a repair/
flaws.
replacement activity.
-3223 Acceptance by Engineering Evaluation Piping items whose examination and evaluation re-
-3000 ACCEPTABLE STANDARDS sults reveal that the criteria of -3221 are not satisfied
-3100 Preservice Examination may be accepted for continued service by engineering evaluation.
Piping items examined prior to commercial service (a) For Class I piping items, this evaluation shall are acceptable for service when the measured wall be conducted in accordance with evaluation methods thickness meets the requirements of the Construction and criteria developed by the Owner.
Code. (b) For Classes 2 and 3 piping items, an acceptable evaluation method and criteria are provided in -3600.
Alternative evaluation methods and criteria may be specified by the Owner.
.3200 Inservice Examination
-3210 General .;3224 Acceptance by Reduction of Time to Next Upon completion of pipe wall thickness examinations, Examination the predicted remaining wall thickness, t,, at the time Piping items whose examination and evaluation re-of the next scheduled examination shall be calculated sults reveal that the criteria of -3221 are not satisfied, for piping items undcr evaluation. The predicted re- are acceptable for continued service when the time to maining wall thickness is the spatial distribution of the next examination for the affected piping items is wall thickness remaining throughout the piping item reduced such that the acceptance criteria of -3221 or and may have a unique value at any given location -3223 are met using the r, for the reduced examination on the piping item. Alternatively, the minimum predicted period.
985 SUPP. 2- NC E4-3
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597-1 CASES OF ASME DOILER AND PRESSM VESSEL CODE FIG. .3220-1 ACCEPTANCE FLOW CHART suPP. 2 - NC 986 E4-4
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Required Thickness Piping item Reference Straight pipe -36001(l1t)
Elbows -3500(al(1) 1 Reducers l -3500(aC(2)
Teesi -3500(a.(3)
Branch -3500a()t3) connections1 Designed item -3500(a)(4)
Other items -3GOOlbl t
Altamate of -3500a1(51 may be used.
Thickness Limit Code class Relerence 1 -3500(cl 2 -3500(d) 3 -3500(e12 2
Alternate criteria may be developed in accordance with -3500M.
FIG. -3500.1 WALL THICKNESS ACCEPTANCE STANDARD FLOW CHART
-3500 Wall ThIckness Acceptance Standards (2) For the small end of concentric and eccentric A flow chart for the acceptance standards is shown reducers. x, shall be not less than 0.875 tnsm for the in Fig. -3500-1. pipe size at the small end. For the large end, the large (a) A Class 1. 2, or 3 butt welded pipe, elbow, end transition and the conical portion, t, shall not be branch connection. or reducer piping item is acceptable less than 0.875 I... for the pipe size at the large end.
for continued service without further evaluation when For the small end transition, the required thickness r, at all locations on the piping item meets the following shall be gradually reduced from that required at the requirements. large end to that required at the small end (see Fig.
(1) For straight pipe and elbows purchased to a *3622-1).
nominal pipe specification with an allowable wall thick- (3) For tees and branch connections, r., shall be ness undertolerance of 12.5%, rt shall be not less than not less than 0.875 r, for the same size pipe for 0.875 tl, except that, for Class I sbort radius elbows, regions outside the limits of reinforcement required by an evaluation shall be conducted to show that the the Construction Code used in the evaluation. For requirements of NB-3642.2 are niet. regions within the limits of reinforcement, lrpshall be 987 E4-5
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597 CASES OF ASME BOILER AND PRESSUtE VESSEL CODE not less than the thickness required to meet the branch (b) Analytical evaluations shall be conducted using reinforcement requirements of the Construction Code. the predicted wall thickness, re,,at the next examination (4) For regions of piping items designed to'specific of the piping item. The methods used to determine tp wall thickness requirements, including designed weld are the responsibility of the Owner.
counterborcs and regions with integral reinforcement, (c) A piping item is acceptable for continued service t, shall be not less than the minimum design thickness, if the minimum pipe wall thickness, branch reinforce-including tolerances and excluding any corrosion allow- *ment requirements, and piping stress criteria of the ances. specified in the original design analysis for the Construction Code used in the evaluation are met for piping item. all specified loading conditions.
(5) As an alternative to the requirements of (d) As an alternative to -3610(c), butt welded pipe,
.3500(a)(2) and -3500(a)(3). for reducers, teed. or branch elbow. branch connection, and reducer piping items connections purchased to fitting standards allowed in may be evaluated In accordance with -3620.
Table NB-3132-1 and for which baseline as-installed (e) Alternative evaluation of pumps, valves, flanges, thickness measurements exist. 1, shall not be less than and other piping items are the responsibility of the 0.875 times the as-installed thickness measurements, Owner.
except that the thickness shall not be less than 0.875 tg.. (19 Piping items under evaluation with ti, exceeding (b) Acceptance criteria for Class 1, 2, and 3 pumps, the acceptance standards of -3500 and satisfying -3600 valves, flanges, reducing elbows, socket weld fittings. shall be monitored for continued degradation. The fre-and any other piping items not covered by -3500(a) quency and means of monitoring for degradation are shall be the responsibility of the Owner. the responsibility of the Owner.
(c) For any Class I piping item, when t, at any location is less than 0.3 rt,,, further evaluation is beyond the scope of this Case. -3620 Evaluation or Pipe, Elbows, Branch (d) For any Class 2 piping item, when :, at any Connections, and Reducers location is less thaii 0.2 t om. further cvaluution is -3621 General Requirements beyond the scope of this Case.
(e) Except as provided in (J) below, for any Class (a) The evaluation shall meet the requirements of 3 piping item, when :p at any location is less than 0.2 -3622 and -3623.
t,,, or 0.5 t ;,, whichever is less, further evaluation (b) For a branch connection or tee, the region within is beyond the scope of this Case. The value of t:,; the limits of reinforcement defined in the Construction shall be determined in accordance with -3600. Code shall meet the requirements of -3624.
(J) As an alternative to -3500(e), decreased wall (c) Evaluations shall be conducted using the appro-thickness, including local through-wall leakage in Class priate piping equations, loadings, load combinations, 3 piping items whose maximum operating temperature allowable material properties, and other acceptance does not exceed 200'F and whose maximum operating standards from the Construction Code used in the pressure does not exceed 275 psi may be accepted. evaluation, except as specifically modified by this Case.
Evaluation methods and acceptance criteria shall be (d) WVben the ratio Rh:, is greater than 50. the specified by the Owner. potential for buckling of the thinned region shall be evaluated. Evaluation methods and acceptance criteria shall be specified by the Owner.
-3600 Analytical Evaluation for Class 2 and Class 3 Piping Items -3622 Thickness Evaluation
-3610 General Requirements -3622.1 Evaluation for Minimum Wall Thickness (a) Analytical evaluations shall be conducted in ac- (a) Except as provided in -3622.1(b). the value of cordance with Construction Code. Later Code Editions rp at any location shall not be less than 90% of the and Addenda may be used. Use of later Code Editions minimum wall thickness of the piping item, 1 mb' required and Addenda shall be reviewed for acceptability to the for design pressure, defined in the Construction Code regulatory and enforcement authorities having jurisdic- used in the evaluation, exclusive of any additional tion at the plant site. corrosion allowance.
QRR E4-6
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE (1) For straight pipe, bends, and elbows, t,,mjshall thinned regions, where )RM is the mean radius of the be determined by: piping item based on nominal wall thickness and L.,,, is the average of the extent of Lm below tR, for the adjacent PD, areas (see Fig. -3622-4). Alternatively, the adjacent 2(S +yP) thinned regions shall be considered a single thinned region in the evaluation.
(2) For concentric and eccentric reducers, tnj, at (b) Provided that the transverse extent of wall thin-each end shall be equal to tr,, of straight pipe of the ning predicted to be less than tmL,, 4L(t) is less than same nominal size as the reducer end. For the conical or equal toJ i, the allowable local thickness, I.,.,
portion of the reducer and the transition at the large shall be determined from Table -3622-1, where Rj.
diamctcr end, r:,n shall be that of the large diameter is the mean radius of the piping item based on the end. A gradual transition in rmin shall be assumed for minimum wall thickness rn,.. For straight pipe, Table the transition at the small end (see Fig. -3622-1). -3622-1 may be-used when L,(,) exceeds v (3) For branch connections and tees, except at except that an additional thickness rb shall be added regions providing reinforcement of the opening required to the value determined from Table -3622-1.
by the Construction Code used in the evaluation, w., (c) This approach shall not be used to evaluate a shall be as required for straight pipe. reducer.
(b) When r, is less than 0.9 ti,,n at any location, additional evaluations may be conducted to determine -36223 Local Thinning- Limited Axial and the allowable local thickness, root, subject to the limita- Transverse Extent tions in (c). The thinned region and the parameters that define the depth and extent of thinning are illustrated (a) When the maximum extent of wall thinning, 4,,
in Fig. -3622-2. The allowable local thickness shall be for which thickness is predicted to be less than tarn is determined in accordance with any one of the methods less than or equal to 2.658/iRi;.dn. and r,,t is greater in -3622.2, -3622.3, -3622.4. -3622.5, or -3622.6. than 1.13 zn,, ta, shall be determined, by satisfying (c) Local thinning evaluation shall not be allowed (b) below and (c) or (d) below. This approach requires for the following: that adequate reinforcement be available surrounding (1) A region adjacent to any branch connection the thinned area in accordance with (c) or (d). This on the run piping, unless the distance between the evaluation approach is not applicable for the following center of the branch connection and the edge of the conditions:
thinned area predicted to be less than t min exceeds Di, (I) Thinned areas adjacent to branch connections, where Di is the nominal inside diameter of the branch when. the reinforcement zone for the thinned area connection and Lm, is the maximum dimension of the would overlap the required reinforcement of the branch thinned region less than 4m;, connection.
(2) At the smaU end transition of a reducer. (2) Thinned areas for which any portion of the (3) Inner portion of elbows and pipe bends (Fig. reinforcement zone would lie on the conical or small
-3622-3), excluding a region within 1.5 R-om?,om of diameter transition zone of a reducer.
the butt welds, unless the 'mmi in the evaluation of (3) Adjacent thinned areas qualified by this ap-
-3622.2, -3622.3, or -3622.4 is replaced by r de- proach when the reinforcement zones associated with fined by: each area would overlap.
(b) The thickness of the remaining pipe wall at the thinned section is adequate if the following equation (05 1 +RhIR cosP.)tm-PPa is satisfied.
-3622.2 Local Thinning - Limited Transverse ta 03534, Extent (a) The evaluation procedure shall consider the depth and extent of the affected area and require that the wall thickness exceed rm,, for a distance that is the (c) If there is a surrounding reinforcement zone with greater of 2.5 IRnom-j.o or 21,,, between adjacent predicted thickness of at least roo for a minimum 989 SUPP. 2 - NC E4-7
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597-1 CASES OF ASME HOULE" AND rESSURE VESSEL CODE dimension of LI2 in all directions, reinforcement for r.i 2,d./D1 the thinned area shall satisfy the following equation. tI. I CoSa (b) For the flared transition at the small end of a concentric reducer, the local allowable thickness shall
,;X 2 (_ L ) (God _ 1) be gradually reduced from the value determined at the conical end of the flare to t,,on for the small end of (d) As an alternative to (c), the reinforcement adjacent the reducer.
to the thinned area shall justify the following equation. (c) This approach shall not be used to evaluate eccentric reducers.
rew 21- (0.935A,:,\
tZh ~m )
-3623 Piping Stress Evaluation
.3623.1 Evaluation Requirements
-3622.4 Local Thinning -UUnlimited Transverse Extent (a) The effects of piping stresses shal be evaluated in accordance with the equations of the Construction (a) The evaluation shall include consideration of the Code used in the evaluation. If the piping design depth and extent of the affected area less than .,,m. analysis is based on nominal piping thickness, the The wall thickness shall exceed trn,, for an axial distance allowable stresses used in a stress analysis based on the greater of 2.5JRr a},rom or 2 L between predicted thickness, t. (see -3210), may be multiplied adjacent thinned regions at each circumferential location by 1.143. Consideration shall be given to changes in on the piping item (see Fig. -3622-5). Alternatively. the pipe metal area, pipe inside area, section modulus.
the adjacent thinned regions shall be considered a single and stress indices or stress intensification factors, as thinned region in the evaluation. described in -3623.2, -3623.3 and -3623.4. The effects (b) Thickness r,], shall be determined from Table of cyclic operating conditions shall be addressed in
-3622-1. accordance with -3625.
(c) This approach shall not be used to evaluate a (b) The piping stress evaluation, shill be based on reducer. the predicted thickness at. each cross section of the
-3622.5 Local Thinning - Elbows and Bent Pipe piping item that exhibits significant thinning or is affected by a change in stress index or stress intensifica-(a) For locations farther than>R'j:tj,, from welds tion factor. Alternatively, the evaluation may be based to adjacent piping items, the predicted thickness on on the limiting cross section.
the outer portion of an elbow or bend may be less than r,,, for straight pipe. The local allowable thickness -3623.2 Nominal Longitudinal Pressure Stresses at each location shall be determined by: (a) The pipe metal area and the pipe inside area, for the thinned cross section might result in stresses
-°: 05 0.5 different from those of the piping stress analysis of toga pea~ n Oe record.
(b) For simplified analysis, the piping item may be (O.)
{R6 assumed to be uniformly thinned with a thickness of t For this approach, the nominal longitudinal pres-where sure stress shall be determined by:
RbIR,,.= ratio of elbow bend radius to mean pipe radius, based on ton for the same size pipe S, PD.
.3622.6 Local Thinning - Central Portions of Concentric Reducers When evaluating reducers, the large and small ends shall be evaluated separately. For the large end, p..in (a) For the conical portion of concentric reducers, shall be determined from all locations for the large the local allowable thickness less than kIn shall satisfy end and conical section. For the small end, i,,,t, for the following equation: the entire reducer shall be used.
SUPP.2 - NC 990 E4-8
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597 CASES OF AME BOILER AND BrSsURENVESSEL CODE (c) Detailed stress analysis may be conducted based (c) Detailed stress analysis may be conducted based on the complete set of measurements around the thinned on a complete set of measurements around the thinned cross-section of the piping item. The nominal longitudi- cross section of the piping item.
nal pressure stress, S,,, sball be determined by: (d) When evaluating thinning at the cross section of a branch connection, the requirements of -3623.2(c)(1) shall be met.
Sp = PA,;
.3623.4 Stress Intensification Factors-and Stress Indices The local piping item wall thickness could affect (I) To evaluate piping at a branch connection the stress indices or stress intensification factors used beyond the limits of reinforcement, it shall be assumed in determination of the effective piping stress at a that the entire region within limits of reinforcement is branch connection. When reduced wall thickness could at thickness tm;n, for the unreinforced pipe section, with increase these factors, the effect shall be considered the outside surface at the pipe nominal outside radius. by using a reduced piping item thickness determined If excess reinforcement is available within the limits in accordance with (a), (b), or (c).
of reinforcement, the excess metal area may be included (a) Except as allowed in (b) or (c), stress intensifica-in A,,. tion factors or stress indices for a piping item shall (2) WVhen evaluating the longitudinal pressure be based on the assumption of uniform wall thickness, stress in the central cone of a reducer, the stress shall using a value of t,,',,, and an associated mean pipe be determined based on the local radius at the cross radius in the formula for these factors.
section and the local t. at and adjacent to the cross (b) As an alternative (a) above, the factors may be section of interest, except that the resulting stress shall based on the average t. of the piping item excluding be multiplied by a factor of l/cosot. branch reinforcement zones, except that predicted thick-(d) When using Code Editions and Addenda that ness at locations within a distance of twice the pipe require use of stress indices, the nominal longitudinal nominal wall thickness from butt welds to adjacent stress determined in accordance with (b) and (c) shall components need not be considered. For reducers, the be doubled. average t, of the small end shall be used with the small end diameter to determine the factor.
-3623.3 Nominal Longitudinal Bending Stresses (c) As an alternative to (a) or (b) above, stress analysis of thinned piping items may be conducted to (a) Thinning of the piping item cross-sectional area show the effects of wall thinning and the distribution might result in bending stresses different from those of stresses on an affected piping item.
of the piping stress analysis of record. The nominal longitudinal bending stress, Sb, for the various loading -3624 Evaluation of Branch Connections conditions and load combinations shall be deter- -3624.1 The region of branch connections and tees mined by: within limits of reinforcement of the Construction Code used in the evaluation shall be evaluated in accordance with -3624.2 or -3624.3.
S Ar,
' + PAJ Z.;n -3624.2 Branch Connections Not Requiring Reinforcement (a) .The region on the piping run shall be evaluated (b) For simplified analysis, the piping item section in accordance with the requirements of -3622 and -3623, modulus may be based on a uniformly thinned section without consideration of the branch connection, except with thickness When W evaluating reducers, the that t4, within a region of radius of Di of the branch large and small ends shall be evaluated separately. For pipe from the center of the branch connection shall the large end, r. ,a,, shall be determined from all locations not be less than :,n for the pipe run.
for the large end and conical section. For the small (b) The branch piping shall be evaluated in accord-end, tain for the entire reducer shall be used. ance with the requirements of -3622 and -3623.
991 E4-9
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597 CASES OF ASME BOILER AND PRESSlRE %ESSEL CODE
-36243 Branch Connections Requiring A,=predicted metal cross-sectional area of Reinforcement pipe, in.2 Ad,, the reinforcement area available in the pipe (a) Branch reinforcement requirements shall be deter-wall based on the predicted thickness distri-mined in accordance with the Construction Code used bution in excess of t.r1 and within the limits in the evaluation.
of reinforcement of the Construction Code (b) For the region of the piping run that provides for an opening with diameter L, at the branch reinforccment. the value of r, at any location region of local thinning. in.2 shall not be less than tma for the nominal pipe run D.=nominal outside diameter of piping item plus any required reinforcement at that location.
(e.g., 10.75 for NPS 10 pipe), in.
(c) For the region of the branch pipe that provides d0 =maximum outside diameter of a reducer at branch reinforcement, to,shall not be less than tin, for the thinned location, in.
the branch pipe plus any required reinforcement.
Di=outside diameter at the large end of the
-3625 Evaluation for Cyclic Operation reducer, in.
Di= nominal inside diameter of a branch connec-(a) For piping items with timin not less than 0.75 tion, in.
rth, and subject to no more than 150 equivalent full f=stress range reduction factor temperature cycles at the time of the next examination. i=strcss intensification factor of the Construc-in accordance with the Construction Code used in the tion Code (not less than 1.0) evaluationpiping stress equations that include thermal l,,=predicted minimum moment of inertia of expansion and anchor movement stresses need not be the thinned pipe about the neutral axis of evaluated. the pipe section, considering all orientations (b) For piping items not meeting the requirements of the section neutral axis, in. 4 of -3625(a), when the design includes consideration of L=maximum extent of a local thinned area thermal expansion stresses, the allowable stress range with wall thickness less than I.,., in.
for expansion stress shall be determined in accordance L, =maximum extent of a local thinned area with the Construction Code used in the evaluation, with wall thickness less than t mn, in.
except that the stress intensification factor. i, shall be 4,(0 )=maximurn axial extent of a local thinned revised to take into account the geometry of the thinned area with wall thickness less than tmi. in.
region. As an alternative to establishing a revised stress L,,.,=nmaximum of the axial extents of two adja-intensification factor, the stress range reduction factors cent local thinned areas with wall thickness of Table -3625-1. which are based on an increase in the stress intensification factor by a factor of 2 over less than tn,in, in.
the life of the component, may be used. L(,,)=,maximum transverse extent of a local (c) The potential for local overstrain in the thinned thinned area with wall thickness less than region for the combination of maximum sustained plus ti,,, in.
thermal expansion stresses shall be considered. Sus- LAvglaverage of the extents of thickness less than tained loads include pressure, weight, and other sus- t,.i. for two adjacent thinned areas, in.
tained mechanical loads. Local overstrain is defined AMb=resulting bending moment from the design in NC-3672.6(b). Evaluation methods and acceptance analysis of record for each loading condition criteria shall be specified by the Owner. under consideration, in-lb P-design pressure, psi
-3626 Nomenclature Rb =bend radius of an elbow to the elbow center A.=total cross-sectional area of pipe based on line, in.
R.=nominal outside radius (e.g., 2.25 for NPS rD2 4 pipe), in.
nominal outside diameter, 4°, 2 Rm,. radius to the nominal outside surface of the Ai=predicted inside cross-sectional area for a pipe plus the nominal distance between the pipe that has experienced wall thinning, in.2 center of the pipe and the neutral axis. in.
A.,=predicted metal cross-sectional area for a Ri.=mean radius of piping item based on the pipe that has experienced wall thinning, iD? nominal outside radius and the minimum 992 E4-10
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597-1 CASES OF ASAE BOILER AND PRESSURE VESSEL CODE Central conical section a
GENERAL NOTE:
Transition zones extend from the point on the ends where the diameter begins to change to the point on the central cone where the cone angle Is constant.
FIG. *3622-1 ZONES OF REDUCER wall thickness (e.g., 7.85 for NPS 16 pipe piping item. For items designed to specified witi t:,,,n= 0.30 in.), in. minimum thickness, the nominal thickness R*,,,=mean radius of piping item based on the is the design thickness, including corrosion nominal radius and thickness (e.g., 6.75 for allowance and excluding tolerances, in.
NPS 14 XS pipe with r,,, = 0.5 in.), in. t,=distribution of predicted local thickness of S=allowable stress for piping item, including a piping item at the next scheduled examina-joint efficiency factor. E, if applicable, psi. tion, in.
S,=maximuni nominal bending stress at the topn = minimum predicted local thickness of a thinned section, psi. piping item at the next scheduled examina-S,=norminal longitudinal pressure stress, psi. lion, in.
to =allowablc local thickness, in. y=factor required by the Construction Code tfr=uniform thickness, of piping item, required used in the evaluation by the Construction Code, to withstand sus- Z,,n=predicted minimum section modulus for the tained and occasional bending loadings in thinned section, Including consideration of the absence of pressure, thermal expansion. the shift of the neutral axis of the thinned and anchor movement loadings, in. pipe section, I0 ,j1 /R., in.3 rtii=m.nimurn wall thickness required by the a=maximumn cone angle at the center of a Construction Code to sustain pressure, ex- reducer, degree clusive of* tolerances and any allowances $=maximum angle from the center of the outer for corrosion, in. one-half of the elbow to the location of the t for. large end of a reducer, in. thinned area being evaluated, as measured t npitcn for straight pipe, in. in the pipe cross section, degree I'min =adjusted minimum thickness for inner por- 6=nominal distance between the center of the tion of an elbow, in. pipe and the neutral axis of the thinned r:. =nominal thickness of pipe or fitting specified piping section, in.
in the applicable industry standard for the 993 SUPP. 2 - NC E4-11
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE InoI Axial dirsetion I Transverse I (hoop direction)
FIG. -3622.2 ILLUSTRATION OF WALL THINNING SUPP. 2 - NC 994 E4-12
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued N-597 CASES OF ASME BOILR AND PRESSURES SSEL CODE FIG. -3622-3 ELBOW AND NOMENCLATURE 995 E4-13
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597 CASES OF ASME BOHIR AND PRESSURPE VESSEL CODE I k t.in in surrounding area Area 3 tp,3< tmin XU . minimum distance between areas iand I
[m - maximum extent of thinned areai Lmavg - 0.5 Lmr. L+m, GENERAL NOTE:
Combination of adjacent areas into an equivalent single area shall be based on dimensions and extents prior to combination.
FIG. -3622-4 SEPARATION REQUIREMENTS FOR ADJACENT THINNED AREAS 996 E4-14
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued N-59i CASES OF ASIE BOIHLR AND PRESSURE ESEL CODE
[Note (1)]
Xi/
- minimum distance between areassandjat anycircumferential location on pipe Lm(a) > - maximum extent of thinned srea in axial direction L sax- maximum of the extents Lm(l S andLm)o two adjacent areas NOTES:
- 11) Areas need not be combined into single areas based on separation in the transverse direction, provided that transverse extents of lndrvidual adjacent thinned areas do not overlap.
121Combination of adjacent areas into an equivalent single area shall be based on dimensions and extents prior to any combination of adjacent areas.
FIG. -3622-5 SEPARATION REQUIREMENTS FOR ADJACENT THINNED AREAS 997 E4-15
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597 CASFS OF ASME DOILER AND PRESSURE VESSEL CODE TABLE-3622-1 Allowable Local Thickness Lt ortstalx4.^a 4 Rit. s 32.2 1 '3622.4 0 0.100 0.100 0.20 0.100 0.261 0.23 0.100 0.300 0.26 0.100 0.375 0.32 0.100 0.477 0.38 0.100 0.551 0.45 0.10D 0.616 030 0.100 0.651 0.60 0.100 0.703 0.70 0.182 0.742 0.83 0.300 0.778 0.05 0.315 0.78z 0.90 0.349 0.794 1.00 0.410 0.813 1.20 0.505 0.841 1.40 0.572 0.860 1.60 0.622 0.873 1.80 0.659 0.883 2.00 0.687 0.891 2.25 0.714 0.897 2.50 0.734 0.900 2.75 0.750 0.900 3.00 0.763 0.900 3.50 0.787 0.900 4.00 0.811 0.900 4.50 0.834 0.900 5.00 0.858 0.900 5.50 0.882 0.900 6.00 0.900 0.900
>6.00 0.900 0.900 GENERAL NOTE:
Interpolation may be usedfor intermediate values.
- 998 E4-16
ENCLOSURE 4 WATTS BAR NUCLEAR PLANT UNIT 1 REQUEST FOR REVIEW AND APPROVAL FOR APPLICATION OF CODE CASE N-597-1 CASE (continued)
N-597 CASES OF ASME BOILER AND PRESSURE VESSEL CODE TABLE -3625-1 MODIFIED STRESS RANGE REDUCTION FACTORS Number of Equivalent Stress Range 2
Full Reduction Factor , f Temperature Cycles', N b50 or less 1.0
>650 to 1100 0.9
>1100 to 20C0 0.8
>2000 to 3900 0.7
>3900 to 8500 0.6
>8500 to 21,000 - 5 over 21,000 0.4 NOTES:
(1) Cyclesto next scheduled Inspection or repairreplacementactivlty.
(2) The modified stress range reduction factors are based on an increase In the stress intensification factor, I, by a factor of 2 over the life cl the component.
999 E4-17