PNP 2014-015, Relief Request Number RR 4-18 - Proposed Alternative Use of Alternate ASME Code Case N-770-1 Baseline Examination
| ML14056A533 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/25/2014 |
| From: | Vitale A Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PNP 2014-015 | |
| Download: ML14056A533 (31) | |
Text
{{#Wiki_filter:Entergy Entergy Nuclear Nuclear Operations, Operations, Inc. Inc. Palisades Palisades Nuclear Nuclear Plant Plant liritei 27780 27780 Blue Blue Star Star Memorial Memorial Highway Highway Covert, MI Covert, Ml 49043-9530 49043-9530 Tel Tel 269 269 764 764 2000 2000 Anthony J. Anthony J. Vitale Vitale Site Site Vice Vice President President PNP 2014-015 PNP February 25, 2014 February U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Relief Request Number RR 4 Proposed Alternative, Use of Alternate ASME Code Case N-770-1 Baseline Examination Palisades Nuclear Plant Docket 50-255 License No. DPR-20
Dear Sir or Madam:
Pursuant to 10 CFR 50.55a(a)(3)(ii), Entergy Nuclear Operations, Inc. (ENO) hereby requests NRC approval of the Request for Relief for a Proposed Alternative for the Palisades Nuclear Plant (PNP). This alternative is for the current fourth 10-year 10-year lSIISI interval. The request is associated with the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Code Case N-770-1, as conditioned by 10 10 CFR 50.55a(g)(6)(ii)(F)(1) and 10 10 CFR 50.55a(g)(6)(ii)(F)(3), dated June 21, 2011. To support the startup startup of PNP following the current refueling outage, ENO requests approval of this alternative by March 8, 8, 2014. This submittal contains no no proprietary information. information. Summary of Summary of Commitments Commitments This This letter letter identifies identifies one one new new commitment, commitment, asas described described in in Attachment 2, 2, and and no no revised revised commitments. commitments.
PNP 2014-015 PNP 201 4-015 Page 22 Page Sincerely, Sincerely,
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Attachments: Relief Request
- 1. Relief Request Number RR RR 4-18 Proposed Alternative
- 2. Description of of Commitment
- 3. Structural Integrity Associates, Inc. Memorandum cc: Administrator, Region III, Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC
ATTACHMENT 11 ATTACHMENT ENTERGY NUCLEAR ENTERGY NUCLEAR OPERATIONS, OPERATIONS, INC. INC. PALISADES NUCLEAR PALISADES NUCLEAR PLANT PLANT RELIEF REQUEST RELIEF REQUEST NUMBER NUMBER RR RR 4-18 PROPOSED PROPOSED ALTERNATIVE in Accordance with 10 CFR 50.55a(a)(3)(ii) in Hardship or Unusual Difficulty Difficulty Without Compensating Increase in Level of of Quality and and Safety Component(s Affected II Applicable Code Edition
- 1. ASME Code Component(s)
Components // Numbers: See Enclosure Table 11 Pressure Retaining Dissimilar Metal Piping Butt Welds 82/182 Containing Alloy 821182 Code of Record: American Society of Mechanical Engineers (ASME) Section XI, 2001 Edition through 2003 Addenda as amended by 10 CFR 50.55a ASME Code Case N-770-1, "Alternative Alternative Examination Requirements and Acceptance Standards for Class 1 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182W861 82 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section Xl, XI, Division 1" 1 N-770-1 Inspection Item: A-2 and B
Description:
Class 11 Pressurized Water Reactor (PWR) pressure retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds containing Alloy 821182 82/182 Inspection Interval: Unit / Inspection Palisades Nuclear Plant (PNP) / Fourth 10-Year Interval December 13, 13, 2006 through December 12, 12, 2015
- 2. Applicable Code Requirements The ASME Boiler and Pressure Pressure Vessel Code, Rules for lnservice Inservice Inspection of Nuclear Nuclear Power Plant Components, Section Xl, XI, 2001 Edition through 2003 Addenda, as amended by 10 10 CFR 50.55a.
With the the issuance of of a revised 10 50.55a in 10 CFR 50.55a in June June 2011, 2011, the Nuclear Regulatory Regulatory Commission Commission (NRC) staff incorporated, incorporated, by by reference, Code Code Case Case N-770-1. N-770-1. Specific Specific implementing requirements are are documented documented in in 10 10 CFR CFR 50.55a(g)(6)(ii)(F) 50.55a(g)(6)(ii)(F) and and are are listed listed below: below: A. Regulation A. Regulation 10 10 CFR CFR 50.55a(g)(6)(ii)(F)(1) 50.55a(g)(6)(ii)(F)(1) states states Licensees "Licensees of of existing, existing, operating operating pressurized water reactors as of July 21, pressurized water reactors as of July 21, 2011 shall 2011 shall implement implement the the requirements requirements of of ASME ASME CodeCode Case Case N-770-1, N-770-1, subject subject to to the the conditions conditions specified specified in in paragraphs paragraphs (g)(6)(ii)(F)(2) (g)(6)(ii)(F)(2) through through (g)(6)(ii)(F)(10) (g)(6)(ii)(F)(10) of of this this section, section, by by the the first first refueling outage outage after August 22, after August 22, 2011. 2011." 11 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE
- 8. Regulation 10 B. Regulation 10 CFR 50.55a(g)(6)(ii)(F)(3) states CFR 50.55a(g)(6)(ii)(F)(3) that baseline states that baseline examinations examinations for for welds in welds Code Case in Code N-770-1, Table Case N-770-1, Table 1, 1, Inspection Inspection Items Items A-1, A-i, A-2, A-2, and and 8,B, shall shall be be completed by completed by the the endend ofof the the next next refueling refueling outage outage after after January January 20, 20, 2012.
2012. The welds The welds covered covered by this proposed by this alternative would proposed alternative would be be classified classified as as Inspection Inspection Items Items A-2 A-2 and B (described and 8 (described below) below) forfor which which visual visual and and essentially essentially 100 100 percent percent volumetric volumetric examination, as examination, as amended amended by 10 CFR by 10 CFR 50.55a(g)(6)(ii)(F)(4), 50.55a(g)(6)(ii)(F)(4), in in part, part, are are required. required. ASME Code ASME Code Case N-770-1, Table Case N-77G-1, Table 1, 1, Examination Examination Categories, Categories, a8 as amended amended by by 10 10 CFR CFR 50.55a(g)(6)(ii)(F) 50.558(g)(6)(II)(F) ~ CLASS 11 PWR CLASS PWR Pressure Pressure Retaining Retaining Dissimilar Dissimilar Metal Piping and Metal Piping and Vessel Nozzle Butt Vessel Nozzle Butt Welds Welds Containing Alloy Containing 82/182 Alloy 821182 Parts Examined Examined Insp Insp Parts Extent and Extent and Frequency Frequency of Examination of Examination Item Item Bare metal Bare metal visual examination each refueling refueling outage. Unmitigated butt weld Unmitigated at Hot Leg Leg operating Essentially 100%100% volumetric examination for axial and temperature (-2410) S A-2 circumferential flaws in accordance with the applicable circumferential 625°F (329°C) requirements of ASME Section XI, Appendix VIII, every five 625°F years. Baseline examinations shall be completed by the end of the next refueling outage after January 20,2012. 20, 2012. Bare metal visual examination once per interval. Unmitigated butt weld Unmitigated at Cold Cold Leg operating Essentially 100% volumetric examination for axial and at temperature (-2410) 2! circumferential flaws in accordance with the applicable circumferential B 525°F (274°C) and < requirements of ASME Section XI, Appendix VIII, every second 580°F (304°C) < inspection period not to exceed 7 years. Baseline examinations shall be completed by the end of the next refueling outage after January 20, 2012. ASME Section Xl, XI, Appendix VIII, Supplement 10, 10, "Qualification Qualification Requirements Requirements for Dissimilar Metal Piping Welds, Welds," is applicable to dissimilar metal (DM) (OM) welds without cast materials.
3. Reason for Request
3. Examination Examinations s of the the DMOM welds listedlisted in Enclosure Table 11 of this request could not be perlormed performed as as required by by ASME ASME Code Case N-770-1, as as conditioned conditioned by by 10 CFR 50.55a(g)(6) (ii)(F). 10 CFR 50.55a(g)(6)(ii)(F). These These DM OM welds welds are are nominal nominal pipe pipe size size (NPS) (NPS) 22 inches inches andand greater greater full full penetration penetration branch branch connection connection welds welds installed installed in in primary primary coolant loop piping. coolant loop piping. SeeSee the the Enclosure Enclosure forfor typical typical configuration configuration.. The The relevant relevant conditions conditions for this request for this request forfor alternative alternative are are ASME ASME Section Section XlXI Code Code Case Case N-770-1, N-770-1, andand 1010 CFR CFR 50.55a(g)(6) (ii)(F) items 50.55a(g)(6)(ii)(F) items (1)(1) and (3), which and (3), which address address performing performing the the required required baseline baseline examinations examinations.. 22 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE Regulation 10 Regulation 10 CFR 50.55a(g)(6)(ii)(F)(1) requires CFR 50.55a(g)(6)(ii)(F)(1) requires thatthat licensees implement the licensees implement the requirements requirements of ASME of ASME CodeCode Case Case N- N-770-1, subject to nO-1, subject to the conditions specified the conditions specified inin paragraphs paragraphs (g)(6)(ii)(F)(2) (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10 through (g)(6)(ii)(F)(1 0)) ofof this this section, section, by the first by the first refueling refueling outage outage after after August August 22, 22, 2011. 2011. Regulation 10 Regulation 10 CFR CFR 50.55a(g)(6)(ii)(F)(3) requiresrequires that baseline baseline examinations examinations for for welds welds inin Code Case Case N-770-1 N-770-1 Table Table 1, 1, Inspection Inspection Items Items A-1,A-i, A-2, and and 8B bebe completed completed by by the the end end of the next refueling next refueling outage outage after January 20,2012.20, 2012. Relief is requested from Relief from 1010 CFR 50.55a(g)(6)(ii)(F) 50.55a(g)(6)(ii)(F) items items (1) and (3) for (1) and performance of for performance required baseline volumetric examinations of the eight cold cold leg leg welds and one hot leg weld listed in the Enclosure Table 1. Hardship The PNP welds in question are outside of the current Performance Demonstration Initiative Initiative (PDI) demonstrated joint configurations. That is, there currently are no POI (POI) PDI demonstrated volumetric techniques for this PNP weld joint configuration. Volumetric examination techniques for this complex weld configuration (see Enclosure Figures 11 and 2) require development and demonstration through the POI PDI qualification program. Under the POI PDI program, representative mock-ups would need to be fabricated in accordance with the POI PDI specimen fabrication program in order to develop examination techniques. Qualification of procedures and personnel would be needed in order to reliably perform qualified examinations of these welds. It is estimated that these activities would take a minimum of approximately 18 months. Attempting "best best effort" effort phased array ultrasonic examinations on these welds for which relief is requested without the needed technique development and demonstrations would not produce reliable results. Even though qualified procedures exist for the weld thickness and diameter, the geometry would negatively impact sound path calibration, search unit focusing, proper inside diameter impingement angles, and cause mis-orientation angles. As a result, this geometric complexity of the configuration would challenge the capability of current procedures to reliably characterize and size indications identified during the examinations and could lead to false-positive indications, as well as unnecessary increased radiation exposure to examination personnel. A design mitigation mitigation strategy has not been been approved for this weld configuration. Creating a mitigation mitigation strategy would require significant significant time for development development of tooling, qualification qualification of of procedures and personnel, and mockup verification, in order to develop an approved method that that would alsoalso minimize radiation exposure to personnel. Due Due to the location to the location of thethe welds, welds, performing performing manualmanual phased phased array ultrasonic ultrasonic test test examinations of of the nine nine welds welds would involve involve significant significant radiation exposure exposure to personnel. personnel. TotalTotal dose dose incurred incurred by by examination, examination, radiation protection, and and supervisory supervisory personnel personnel during during ultrasonic ultrasonic testing testing ofof the nine nine weld weld locations locations is is estimated estimated to to be be at at least 37 rem. This least 37 This total total includes preparation activities, and preparation activities, and credits dose reduction credits dose reduction controls controls andand measures measures such such as as shielding, shielding, decontamination decontamination of of components, components, high high efficiency efficiency particulate particulate airair filter filter ventilation ventilation units, units, cameras, cameras, and remote telemetry. Dose and remote telemetry. Dose rates vary, rates vary, onon contact, contact, from from 120 120 mremlhour mremlhour to to 15000 15000 mrem/hour mrem/hour for the weld for the weld locations. locations. General General areaarea dose dose rates rates inin the the vicinity vicinity of of the weld locations the weld locations vary vary from from 25 25 mrem/hour mrernlhour to 200 mrem/hour. to 200 mrem/hour. Development Development of of aa qualified qualified procedure procedure for for the the volumetric volumetric inspections inspections would would include include 33 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE identification of identification of examination examination techniques techniques thatthat would would minimize minimize radiological radiological exposure exposure to to examination personnel. examination personnel.
- 4. Proposed
- 4. Alternative and Proposed Alternative and Basis Basis forfor Use Use Proposed Alternative Proposed Alternative
- 1) Perform
- 1) Perform periodic periodic system system leakage leakage tests tests inin accordance accordance with with ASME ASME Section Section XI Xl Examination Examination Category B-P, Table IWB-2500-1 Category B-P, Table IWB-2S00-1 (Reference 10). (Reference 10).
- 2) Perform
- 2) Perform visual visual andand dye penetrant surface dye penetrant surface examinations examinations of of the the welds welds in in accordance accordance with with ASME requirements.
ASME requirements. During During the 2012 (1R22) (1R22) andand 2014 (1R23)(1R23) refueling refueling outages, outages, visual and external and external surface examinations examinations of certain welds for which which relief relief is is requested requested identified identified no evidence evidence of of through-wall through-wall cracking or or leakage leakage for these components, components, as identified identified in in Enclosure Table Enclosure Table 1. 1. 50.55a(3)(ii), ENO proposes to perform appropriate actions to meet Pursuant to 10 CFR SO.SSa(3)(ii), ASME Section XI Xl Code Case N-770-1 examination requirements, requirements, as required, for those dissimilar metal welds identified in Enclosure Table 11 of this request during the first refueling outage after a viable technology is developed to perform these examinations. examinations. Basis for Use Enclosure Table 11 describes the eight cold leg welds and one hot leg weld that have not been examined in accordance with Code Case N-770-1 examination requirements, requirements, as required by 10 CFR SO.SSa(g)(6)(ii)(F) 50.55a(g)(6)(ii)(F) items (1) and (3). Examination History Enclosure Table 1 1 also provides examination history information for the nine weld locations for which relief is requested. No evidence of through-wall cracking for these components has been identified during these inspections. Moreover, Moreover, for the three weld locations that were not subject to surface or visual examinations during the ongoing 11R23 refueling outage (i.e., (Le., weld no.s no.'s 3, 6, and 7 in Enclosure Table 1), 1), maintenance activities in the vicinity of the weld locations locations duringduring the 11R23R23 refueling outage did not identify did not identify observations of of leakage leakage from the welds. Structural Structural Evaluation Evaluation Structural Structural Integrity Integrity Associates, Inc. Inc. (SIA) (SIA) performed performed aa bounding bounding evaluation evaluation of of the the Alloy Alloy 82/1 821182 82 full full penetration penetration weld weld which which connects connects the the hot hot leg leg to the drain to the drain nozzle nozzle (see(see SIA. SIA Memorandum Memorandum in in Attachment Attachment 3). 3). The The evaluation evaluation concluded concluded the the following: following:
- Because Because of of the the post post weld weld heat heat treatment treatment (PWHT)
(PWHT) of the nickel-base of the nickel-base materials, materials, there there isis aa low low probability probability that that aa PWSCC PWSCC crack crack of of engineering engineering size size has has initiated initiated on on the the Alloy Alloy 82/182 821182 full-penetrati full-penetrationon branch branch pipe pipe connection connection weldswelds at at PNP. PNP. Engineering Engineering size size isis defined defined as as aa flaw flaw with with aa depth depth ofof about about 11 to to 22 millimeters. millimeters. 44 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE
** The finite-element calculations The finite-element calculations for for the the hot-leg hot-leg drain drain nozzle nozzle show show relatively relatively modest modest peak total peak total tensile tensile stresses stresses on on the the wetted wetted surface surface forfor normal normal operating operating conditions conditions due due to the benefit to the benefit ofof the PWHT applied.
the PWHT applied.
** ASME Code ASME Code acceptance acceptance criteria criteria areare satisfied satisfied for for 60 effective full 60 effective full power years for power years for aa circumferential flaw, circumferential flaw, and and more more than than 34 years for 34 years for an an axial axial flaw flaw assuming assuming crack crack initiates at day initiates one. Using day one. Using hot hot leg leg crack crack growth rate rate and temperature.
temperature. oo PWHT reduced the crack growth growth rate for Alloy 182 182 weld metal (e.g., by a factor of between two and four). This benefit benefit is is conservatively not credited in the the MRP-1 15 crack growth rate equation for Alloy 182. MRP-115 182. o susceptibility to PWSCC initiation is greatly reduced The susceptibility reduced for nickel-based weldments operating at reactor cOld-leg cold-leg temperature, and the PWSCC crack growth rate at reactor cold-leg temperature is approximately four times lower than the corresponding crack growth rate at reactor hot-leg hot-leg temperature
- An additional limit analysis was performed for the hypothetical partial-arc through-wall circumferential flaw illustrated in Figure 10 10 in Attachment 3. This flaw is predicted to be through-wall for 45 degrees in approximately 100 years using the MRP-1 15 crack growth rate equation. In the unlikely case of initiation of a MRP-115 circumferential crack and the unlikely case that a circumferential crack were to grow to a large size, non-axisymmetric crack growth behavior would be expected ultimately to result in detection of leakage prior to the possibility of unstable pipe rupture.
- Another limit analysis was performed in order to investigate the stability of a hypothetical axial flaw that has grown through-wall to encompass the entire Alloy 82/182 weld cross section and a large portion of the Alloy 600 nozzle. The extent of 821182 this conservatively assumed axial flaw is shown in Figure 11 11 in Attachment 3. The analysis, which applied Level A Service Limits of the ASME Code, showed that the flaw remains stable. This limit analysis shows that the structural stability provided by the pipe branch connection geometry would be expected to preclude the possibility of a rupture. Leakage and not rupture would be the ultimate result of growth of an axial flaw. (Note:
(Note: used two times upset pressure 2650 psi.) Operating Operating Conditions The operating temperature of a component is is a primary factor influencing the initiation of PWSCC. Research by the Electric Power Research Institute (EPRI) (Reference (Reference 8) indicates that the difference difference in in the operating temperature temperature between between hot hot leg leg locations locations and cold leg leg locations locations is is sufficient sufficient to to significantly influence influence the the time toto initiation initiation ofof PWSCC, with the the susceptibility increasing increasing with with temperature. The The research reports PWSCC PWSCC is is least least likely likely to occur occur inin cold leg temperature penetrations. All one of All but one of the the welds welds covered covered byby this this relief areare primarily primarily found found in in lower lower temperature temperature regions regions of of the the system, typically at system, typically at temperatures temperatures near near to to Tcold, Tcold, which which is is approximately 537 537 °F. This means, of. This means, for for these these welds, welds, there there is is aa lower lower probability probability of of crack crack initiation, initiation, and and aa slower crack growth slower crack growth rate (Reference rate (Reference 9). 9). 55 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE Leakage Detection Leakage Detection Capabilities Capabilities Even there were Even ifif there were to to be be flaws in in the the welds for which relief for which relief is is requested, requested, and and these these flaws flaws led led leakage, the leak to leakage, leak detection detection methodology methodology presently presently usedused by industry is by industry is very sensitive. sensitive. After a number After number of recent recent operating events, events, the industry industry imposed imposed an an NEI 03-08, "Guideline NEI 03-08, Guideline for the Management of Materials Materials Issues," Issues, requirement requirement to improve improve leakleak detection capability. result, virtually As a result, virtually all pressurized pressurized water reactors reactors (PWRs) (PWRs) in in the United United States, including PNP, have PNP, have a leak detection capability of less less than or equal to 0.1 gpm (Reference 7). All All plants, including PNP, also monitor seven-day seven-day moving averages of reactor coolant system rates. leak rates. Action response times following a detected primary coolant system leak vary, based based on the action level exceeded and whether containment entry is required to identify the source of the leak. Action levels have been standardized for all PWRs, and are based on deviations from:
- the seven day rolling average,
- specific values, and
- the baseline mean.
Leak rate action levels are identified in Pressurized Water Reactor Owners Group (PWROG) WCAP-1 6465, and are stated below: report, WCAP-16465, Each PWR utility is required to implement the following standard action levels for reactor coolant system (RCS) inventory balance in their RCS leakage monitoring program. Action levels on the absolute value of unidentified RCS inventory balance (from surveillance data): Level 1 1 - One seven day rolling average of unidentified RCS inventory balance values greater than 0.1 gpm. Level 22 - Two consecutive unidentified RCS inventory
- inventory balance values greater than 0.15 gpm.
Level 33 - One unidentified
- unidentified RCS inventory inventory balance value greater than 0.3 gpm. gpm.
Note: Note: Calculation of the absolute RCS inventory balance values must include include the rules for the treatment of negative of negative values and missing observations. Action levels levels on the deviation from the baseline mean: mean: Level 11 - Nine
- Nine consecutive consecutive unidentified unidentified RCS RCS inventory inventory balance values values greater greater than than the the baseline mean mean [pJ [J,J] value.
value. Level Level 22 - Two
- Two of of three consecutive consecutive unidentified unidentified RCS RCS inventory inventory balance balance values values greater than greater than [p [J,J ++ 2o],
20], where where a0 is is the the baseline baseline standard standard deviation. deviation. Level - One unidentified Level 33 - One unidentified RCS RCS inventory inventory balance balance valuevalue greater greater than [J,J +3o]. than [p +30]. These These action action levels levels have have been been incorporated incorporated into into PNP PNP procedures. procedures. 66 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE A A small steam leak small steam leak from weld flaw from aa weld flaw would, would, over over time, time, result result in in aa rise rise in in containment containment sump sump level rate level of increase. rate of increase. Containment Containment sump sump level level isis continually continually monitored, monitored, and rise in and ifif aa rise in the the rate of rate of containment containment sump level increase sump level increase is is observed, observed, plant plant procedures procedures directdirect plant plant operators operators toto identify the source identify the source ofof the the leakage. leakage. Operators Operators may may also also be alerted to be alerted to aa leak leak from from aa flaw flaw by by containment radiation containment monitoring instrumentation. radiation monitoring instrumentation. This This instrumentation, instrumentation, required required by by the 3 Technical Specifications, Specifications, is is capable of of detecting detecting aa 100 100 cm/min 3 cm leak in
/min leak in 45 minutes, minutes, basedbased on on 11% failed fuel. Periodic
% failed Periodic system leakage leakage tests are performed performed in in accordance with ASME Section XI.
Section Xl. Operator Operator walkdowns of containment are periodically performed performed during power operations at lower levels levels of containment containment to to detect leakage. detect leakage. Therefore, with the periodic system leakage tests, the visual and surface examinations examinations performed duringduring 2012 and 2014 refueling outages, the results results of the SIA evaluation, and containment monitoring activities, an acceptable level of quality and safety is provided for identifying degradation from PWSCC prior to a safety-significant flaw developing.
- 5. Duration of Proposed Alternative The duration of the proposed alternative is until the first refueling outage after a viable technology is developed to perform these examinations.
- 6. References
- 1. 10 CFR 50.55a, "Codes Codes and standards,"
standards, July 25, 2013.
- 2. ASME Section XI, "Rules Rules For Inservice Inspection of Nuclear Power Plant Components," Components, 2001 Edition with Addenda through 2003.
- 3. ASME Section Xl, XI, Division 1, Code Case N-460, "Alternative Alternative Examination Coverage for Class 11 and Class 2 Welds, Section Xl, XI, Division 1." 1.
- 4. Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-1 (MRP-139),39), Revision 1, 1, EPRI, Palo Alto, CA, 2008 [ADAMS Accession Number ML1 009700671].
ML1009700671].
- 5. Nondestructive Evaluation: Procedure for Manual Phased Array Ultrasonic Examination of Dissimilar Metal Welds, EPRI-DMW-PA-1, EPRI-DMW-PA-1, Revision Revision 3, 1016645, EPRI 3,1016645, EPRI Palo Alto, CA, 2008.
2008.
- 6. Changing "Changing the Frequency of of Inspections for PWSCC Susceptible Susceptible Welds at Cold Leg Temperatures, Temperatures", in Proceedings Proceedings of 2011 ASME Pressure Vessels and Piping Piping Conference, July 17-21, July 2011, Baltimore, MD.
17-21, 2011, MD. 7.
- 7. WCAP-16465-NP, Rev. 0, 0, Pressurized "Pressurized WaterWater Reactor Reactor Owners Group Standard Standard RCS Leakage Action Levels and Response Guidelines Leakage Action Guidelines for for Pressurized Pressurized Water Reactors, Reactors,"
Westinghouse ElectricElectric Co., Co., September 2006 2006 [ML07031 [ML070310082]. 0082].
- 8. Electric Power
- 8. Electric Power Research Research Institute:
Institute: PWSCC PWSCC of of Alloy 600 Materials Alloy 600 Materials in in PWR PWR Primary Primary System Penetrations, System Penetrations, EPRI,EPRI, Palo Palo Alto, Alto, CA, CA, 1994, 1994, TR-1 03696 [MLO131 TR-103696 [ML013110446]. 10446]. 9.
- 9. Materials Materials Reliability Reliability Program Program CrackCrack Growth Growth RatesRates for for Evaluating Evaluating Primary Primary Water Water Stress Stress Corrosion Corrosion Cracking Cracking (PWSCC)
(PWSCC) of of Alloy Alloy 82, 182, and 82,182, and 132 132 Welds Welds (MRP-1 (MRP-115), EPRI, Palo 15), EPRI, Palo Alto, Alto, CA, CA, 2004, 2004, 1006696 1006696 [MLO51 [ML0511100204]. 00204]. 77 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE
- 10. PNP
- 10. PNP Technical Technical Specification Specification Surveillance Surveillance Procedure Procedure RT-71A, RT-71A, "Primary Primary Coolant Coolant System, System, Class 11 System Class System Leakage Leakage Test,"
Test, Revision Revision 18. 18.
- 11. Pressurized
- 11. Pressurized Water Water Reactor Reactor (PWR)
(PWR) Owner's Owners Group Group Letter Letter OG-12-89, OG-1 2-89, "Transmittal Transmittal of of Final Relief Request Famework under
'Final Relief Request Famework' Relief Request under Relief Request for for Large Large Diameter Diameter Cold Cold Leg Leg Locations with Locations with Obstructions Obstructions (PA-MSC-0934),"
(PA-MSC-0934), MarchMarch 8, 8, 2012. 2012. Enclosure
- 7. Enclosure Table 11 -Weld History Weld Examination History Figure 11 - Nozzle Assembly Materials Figure - Materials (Representative)
Figure 2 - Hot Leg Drain Nozzle Configuration (Representative) 88 of of 88
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE ENCLOSURE ENCLOSURE Table 11 Table Weld Examination Weld Examination History History No. No. Description Description ISI Weld ID ISIWeldlD Location Location 1R19 1R19 1R20 1R20 1R21 1R21 1R22 1R22 1R23 1R23 Examinations Examinations Examinations Examinations Examinations Examinations Examinations Examinations Examinations Examinations Visual Visual inch Cold Surface Surface
- 1. 22 inch Leg Cold Leg PCS-30-RCL-1 A-PCS-30-RCL-l P-50A Discharge P-50A Discharge (Report# 4046 (Report# 4046
- 1. Charging Nozzle Nozzle 11/2 Leg (Report#
(Report# 11 R23-R23-Charging 11/2 Leg Exam number Exam number PT-14-025) PT-14-025) 06-26) Visual inch Cold Surface Surface
- 2. 22 inch Leg Cold Leg PCS-30-RCL-1A-PCS-30-RCL-l A- P-50A Suction P-50A Suction (Report# 4047 (Report# 4047
- 2. Drain Nozzle Nozzle (Report# 11 R23-R23-Drain 5/2 Leg Exam number PT-14-031)
PT-i4-031) 07-28.1) inch Cold Leg 33 inch Leg Visual Surface
- 3. Pressurizer Spray PCS-30-RCL-1 B-PCS-30-RCL-l P-50B Discharge
- 3. Pressurizer Spray (Report# VT VT-b- (Report# 11 R22-Nozzle 10/3 Leg Nozzle 069) PT-12-039) inch Cold Leg 2 inch Leg PCS-30-RCL-1 -i B- Visual Surface 2 P-50B Suction
- 4. Drain Nozzle Nozzle (Report# VT-l0-VT-i 0- (Report #1 R23-R23-Drain 5/2 Leg 048) PT-14-032) inch Cold 2 inch Cold Leg Leg PCS-30-RCL-2A- Visual Surface 2 PCS-30-RCL-2A- P-50C Discharge
- 5. Charging Nozzle Nozzle (Report# VT (Report# 11 R23-R23-Charging 11/2 Leg 083')
083) PT-14-019) 3 inch Cold Leg 3 inch Visual Surface
- 6. PCS-30-RCL-2A-PCS-30-RCL-2A- P-50C Discharge
- 6. Pressurizer Spray Pressurizer Spray (Report# VT (Report# 11 R22-11/3 11/3 Leg Nozzle Nozzle 035) PT-12-032) inch Cold Leg 22 inch Leg PCS-30-RCL -2A- P-50C Suction Visual PCS-30-RCL-2A-
- 7. Drain (Report# VT Drain Nozzle Nozzle 5/2 5/2 Leg 038) 038) 22 inch inch Cold Leg Leg Visual Surface Surface
- 8. Drain PCS-30-RCL -2B-PCS-30-RCL-2B- P-50D Suction Suction
- 8. Drain and and 5/2 Leg VT-l0-(Report# VT-i 0- (Report# 11R23-Letdown Letdown Nozzle Nozzle 071) 071) PT-14-020)
PT-i4-020) Visual Visual Visual Visual Visual Visual 22 inch Visual Visual Visual Visual Hot Leg inch Hot Leg PCS-42-RCL-i -1 H-H- A (Report# (Report# 4047 4047 (Report# VT (Report# VT (Report# 11R23-(Report# R23-
- 9. Drain A Hot Hot Leg Leg (Report# VT-i (Report# VT-l0-0- (Report# 11R22-(Report# R22-Drain Nozzle Nozzle 3/2 3/2 Exam Exam number number 062) 062) VT-14-059)
VT-i 4-059) 022) 022) VT-12-076) VT-12-076) 07-23.1) 07-23.1)
PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE ENCLOSURE ENCLOSURE Figure 11 Figure Nozzle Assembly Materials Nozzle Materials Alloy 821182 Stainless Steel Cladding lCold 4----i Carbon Steel Cold Leg Full Penetration DM Alloy 821182 82/182 Weld Alloy 600 Nozzle Not to Scale Scale
PROPOSED ALTERNATIVE ENCLOSURE Figure 2 Hot Leg Drain Nozzle Configuration (Representative) (excerpt from PNP vendor drawing VEN-M1-D Sheet 108, Revision 8) I
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ATTACHMENT 22 DESCRIPTION OF DESCRIPTION OF COMMITMENT COMMITMENT identifies actions discussed in this This table identifies this letter letter for which Entergy Entergy Nuclear Nuclear Operations, Inc. (ENO) commits to perform. Any other actions discussed in this this submittal are described for information information only only and are are not commitments. TYPE SCHEDULED (Check one) COMPLETION COMMITMENT ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required) ENO will perform appropriate actions to meet V
./ The first ASME Section XI Xl Code Case N-770-1 refueling outage examination requirements, as required, for after a viable those dissimilar metal welds identified in technology is , Enclosure Table 1, of this developed to request during the first refueling outage after perform these a viable technology is developed to perform examinations.
examinations. these examinations.
ATTACHMENT 33 ATTACHMENT STRUCTURAL INTEGRITY STRUCTURAL INTEGRITY ASSOCIATES, ASSOCIATES, INC. INC. MEMORANDUM MEMORANDUM Nuclear Plant Hot Leg Drain Evaluation of the Palisades Nuclear Drain Nozzle for Primary Water Stress Corrosion Cracking February 25, 2014 RAM-i 4-008 RAM-14-008 16 16 Pages Pages Follow Follow
S) Structural Structural Integrity Associates, Inc. 5215 5215 Hellyer Ave. HellyerAve. Suite Suite 210 210 San San Jose, Jose, CA CA 95138-1025 95138-1025 Phone: Phone: 408-978-8200 408-978-8200 Fax: Fax: 408-978-8964 408-978-8964 www.structinlcom www.structint.com rmattson@structinlcom rmattson@suctintcom MEMORANDUM MEMORANDUM February 25,2014 25, 2014 RAM-14-008 TO: William Sims FROM: Dick Mattson
SUBJECT:
Evaluation of the Palisades Nuclear Plant Hot Leg Drain Nozzle for Primary Water Stress Corrosion Cracking Structural Integrity Associates has been contracted by Entergy to evaluate the Alloy 82/182 full penetration weld which connects the hot leg to the drain nozzle. The evaluation focuses on the probability of occurrence of primary water stress corrosion cracking (PWSCC), and the PWSCC growth of a postulated axial and circumferential flaw in the weld. This memorandum summarizes the results obtained to date. Finite Element Analyses A three-dimensional finite element model encompassing 90° of the circumference was constructed using the ANSYS software. The weld was modeled with eighty-seven nuggets representing the lumped weld beads connecting the hot leg to the nozzle. Figure 11 depicts the finite element model, and Figures 2 and 33 depict the weld and patch nuggets, respectively. Analyses were performed for the following steps of of construction: 1.
- 1. Deposit cladding cladding on hot hot leg inside surface (ID). (lD).
2.
- 2. Install drain line nozzle/backingnozzlelbacking ring and deposit weld.
3.
- 3. Remove backing backing ring and deposit ID patch. patch.
4.
- 4. Post weld heat heat treatment, treatment, including creep creep effects effects basedbased upon upon experimental experimental data.
5.
- 5. Subject Subject the configuration to aa hydrostatic hydrostatic test.
6.
- 6. Impose Impose five cycles cycles of of shake "shake down down" at at normal operating operating temperature temperature and and pressure.
pressure. Stress Stress results results normalnormal to to aa circumferential circumferential crack crack are are shown shown in in Figure Figure 4, 4, and and those those normal normal to to anan axial axial crack crack are are shown shown in in Figure Figure 5. 5. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ToIl-Free Toll-Free 877-474-7693 877-474-7693 _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ Akron.OH AIIran, 011 Aibaqu.rque,NM
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Mr. William Mr. William SimsSims February 25,2014 February 25, 2014 RAM-14-008 RAM-14-008 Page Page 22 ofof 16 16 Crack tip Crack tip elements elements along along the nozzle-weld boundary were the nozzle-weld were evaluated evaluated at at seven depths through seven depths through the the thickness, and and atat 0°,30°, 00, 30°, 60° 60° and 90° angular and 90° angular locations, locations, forfor aa postulated circumferential circumferential flaw.flaw. A A plot ofof the applied stress the applied stress intensity intensity vs. vs. crack crack depth depth is is shown shown in in Figure 66 for for the the circumferential circumferential flaw. flaw. Likelihood of PWSCC Initiation Assessment of Likelihood The likelihood of of PWSCC initiation occurring occurring on on the wetted Alloy 600 600 and Alloy 821182 82/1 82 surfaces of of the Palisades hot leg drain nozzle was assessed by Dominion Engineering, Inc. (DEl). This assessment concludes that there is (DEI)). is aa low probability that a stress corrosion crack of of engineering size has initiated in the Alloy 82/ 82/182 182 full penetration branch pipe connection welds at Palisades, including in the hot leg drain nozzle and full penetration weld. The low probability of initiation is the result of the nickel-based material being exposed to the post weld heat treatment (PWHT) applied to the adjacent carbon steel material. The greatly reduced susceptibility to the occurrence ofPWSCC of PWSCC of Alloy 600 weldments that have been exposed to PWHT after welding is demonstrated by the following:
- investigations including studies that show a very significant relaxation detailed laboratory investigations of the residual stress in the surface layer of the weldment,
- PWR plant experience showing hundreds of cases of PWSCC when the Alloy 82/182 or Alloy 600 material was not exposed to PWHT, but extremely few cases when the material was exposed to PWHT subsequent to welding, and,
- the favorable operating temperatures of the Alloy 600 branch connection nozzles at Palisades (eight operate at the relatively low reactor cold leg temperature, and only the single hot leg drain nozzle operates at reactor hot leg temperature) temperature)..
Moreover, the finite element analyses for the hot leg drain nozzle show relatively small peak total tensile stresses on the wetted surface for normal operating conditions due to the benefit of the applied PWHT. As discussed in the DEl DEI letter, given the large demonstrated sensitivity of initiation time to surface stress, there is a low probability that PWSCC initiation of a flaw of engineering size has occurred on this weldment at Palisades. Because of the cold leg operating temperature of the nozzles located on the Palisades cold legs, and because the weld residual stress is expected to be similar for the cold leg locations, this conclusion extends to the eight cold leg nozzles. Crack Growth Evaluation Growth of circumferent circumferential ial and axial flaws was was investigated in order to assess assess the consequence consequencess of such such hypothetical cracking the in the unlikely case of PWSCC initiation. Using the applied stress stress intensity factors described above for a circumferentcircumferential ial flaw, crack crack growth through the depth of of the weld was was calculated, and the results and the results are are presented in in Figure Figure 77 for the circumferent circumferential ial flaw. For the axial flaw, a conservative classical fracture mechanics solution was used for an an elliptical flaw which which has has aa constant width width along along the the inside surface surface of thethe Alloy 600 nozzle Alloy 600 nozzle andand Alloy Alloy I Letter Letter from G. White (Dominion G. White (Dominion Engineering, Engineering, Inc.) Inc.) to to W. W. Sims Sims (Entergy), (Entergy), Effect "Effect of ofPost-Weld Post-Weld Heat Heat Treatment Treatment Applied Applied to to Alloy Alloy 82/1 82 Full-Penetration 82/182 Full-Penetration Branch Branch Pipe Pipe Connection Connection Welds at Palisades, Welds at Palisades," L-4199-OO-O1, L-4199-00-01, Rev. Rev. 0, 0, dated dated February February 25, 25, 2014. 2014. l) Structura! Structural Integrity Integrity Associates, Inc.e Associates, Inc.
Mr. William Mr. William Sims Sims February 25,2014 February 25, 2014 RAM-14-008 RAM-14-008 Page Page 33 ofof 16 16 82/182 weld, 82/182 weld, and variable depth and aa variable through the depth through the thickness. thickness. ThreeThree widths widths forfor the the elliptical elliptical flaw flaw were modeled. were modeled. Figures Figures 88 andand 99 present present the applied stress the applied stress intensity intensity factors factors vs. crack depth vs. crack depth and and crack growth crack growth for for the the axial axial flaw, flaw, respectively. respectively. However,However, more more realistic, realistic, less less conservative conservative results results can be can obtained using be obtained finite element using finite element analysis analysis techniques techniques for for the the axial axial flaw flaw asas was was done done for for the the circumferent circumferential flaw. ial flaw. The The crack crack growth growth law law used used in developing these in developing these plots plots isis as as described described in in MRP-1 15. MRP-115. As discussed As discussed in in the the DEI DEl letter, letter, laboratory laboratory crackcrack growth growth testing testing hashas shown significant benefit shown aa significant benefit ofof PWHT in PWHT in reducing reducing the the crack crack growth growth raterate for for Alloy Alloy 182 182 weld weld metal metal (e.g., (e.g., by by aa factor factor of of between between two and two four). This and four). This benefit is is conservatively conservatively not not credited credited in in the the MRP-MRP-1I 15 15 crack crack growth growth rate rate equation for for Alloy 182. 182. Limit Analysis Limit Circumferential Cracking: Circumferential At 60 effective full power years, the finite element model described above was modified to cracking at the four circumferential include "cracking" circumferential locations. A limit analysis, as described in ASME Code, Section III, Subparagraph Subparagraph NB-3228.1NB-3228. 1,, was performed, taking into consideration the effects of the flux Alloy 821182 82/182 weld by decreasing the weld metal effective yield strength by the "Z" Z factor. The limit analysis results satisfy the Section III criteria. In the unlikely event that circumferential circumferential PWSCC were to occur in the Alloy 82/ 182 full 82/182 penetration weld, the significant variability in residual stress with azimuthal position around the nozzle (for the geometry of a nozzle welded into a cylindrical pipe) would tend to drive crack growth through-wall along part of the circumferenc circumference. e. This non-axisymm non-axisymmetric etric crack growth behavior would be expected ultimately to result in detection of leakage prior to the possibility of unstable pipe rupture. rupture. An additional limit analysis was performed for the hypothetical partial-arc through-wall circumferent circumferential ial flaw illustrated in Figure 10. This flaw is predicted to be through-wall for 45° in approximatel approximately 100 years using the MRP-1 y 100 MRP-l 15 crack growth rate equation. The The analysis, which applied Level A Service Limits of the ASME Code, Code, showed that that the flaw remains stable stable at twice the applied applied loads. Axial Cracking: For the axial For the axial flaw, the the flaw modeled modeled in in ANSYS ANSYS was was conservative conservatively assumed to ly assumed to be be at aa depth of of 75% 75% through-wall through-wall.. The The analysis analysis satisfies satisfies the the Section Section III criteria. Per Figure criteria. Per Figure 9,9, this this equates to 34 effective equates to 34 effective full power power years years of of operation. operation. As As noted noted above, above, if if more more detailed detailed finite finite element analyses were element analyses were performed, performed, therethere would would be be an an increase increase in in the the effective effective fullfull power power years years ofof operation operation correspondin corresponding g to to aa flaw flaw depth depth of 75% through-wall of75% through-wall.. Another limit Another limit analysis analysis waswas performed performed in in order order to to investigate investigate the the stability stability of ofaa hypothetical hypothetical axial flaw axial flaw that has grown that has grown through-wall through-wall to to encompass encompass the the entire Alloy 82/182 entire Alloy 821182 weld weld cross cross-section and a large portion section and a large portion of the Alloy 600 of the Alloy 600 nozzle. nozzle. The The extent extent ofofthis this conservative conservatively ly assumed axial assumed axial flawflaw isis shown shown in in Figure Figure 11.1 I. TheThe analysis, analysis, which which applied applied Level Level A A Service Service Limits Limits ofofthethe ASME ASME Code, Code, showed showed that that thethe flaw remains stable flaw remains stable at at 1.87 1.87 times times the the applied applied e StructuralIntegrity StructoraI Integrity Associates, Associates, Inc Inc'"
Mr. William Mr. William Sims Sims February February 25,2014 25, 2014 RAM-14-008 RAM-14-008 Page 44 of Page of 16 16 loads. This loads. This limit limit analysis analysis shows shows that thethe structural structural stability stability provided by the the pipe branch connection geometry connection geometry would be be expected expected to preclude the the possibility ofof aa rupture. Leakage Leakage and not and rupture would be not rupture be the result of the ultimate result growth of of growth of an an axial axial flaw. flaw. Conclusions Conclusions calculations performed, itit is Based on the assessments and calculations is concluded concluded that the visual examinations examinations for evidence of pressure boundary leakage required by ASME Code Case N-722-1 sufficient to ensure are sufficient ensure that nuclear safety is maintained (i.e., periodic volumetric/surface examinations for indications of PWSCC are not necessary). The adequacy of of visual examinations to address the PWSCC concern is demonstrated by the following:
- Because of the PWHT of the nickel-based materials, there is a low low probability that a stress corrosion crack of engineering size has initiated on the Alloy 82/ 182 full 82/182 penetration branch pipe connection welds at Palisades. Confidence in this conclusion is provided by a combination of laboratory investigations, extensive plant experience, the favorable operating temperatures, and finite element weld residual stress analyses specific to Palisades.
- In the unlikely case that crack initiation were to occur, crack growth calculations considering PWSCC as the failure mechanism demonstrate that the hot leg drain nozzle weldment satisfies ASME Code acceptance criteria for 60 effective full power years for a circumferential flaw, and more than 34 years for an axial flaw. These results are highly conservative in that they assume a crack could initiate, that a crack initiates immediately at the start of plant operation, and that a conservative limit load analysis is satisfied. In the unlikely case that a crack has already initiated, it would most likely have occurred closer in time to today than to plant startup.
- The ultimate result of any circumferential or axial cracking would very likely be detection of leakage prior to the possibility of unstable pipe rupture. In the unlikely case of initiation of an axial crack and the unlikely case that an axial crack were to exceed a depth of 75% through-wall, the structural stability provided by the pipe branch of75%
connection geometry would be expected to preclude the possibility of a rupture. Leakage and not rupture would be the ultimate result of growth of an axial flaw. Similarly, Similarly, in the unlikely case of initiation of a circumferential crack and the unlikely case that a circumferential circumferential crack were to exceed a depth of 60% through-wall, non-axisymmetric crack growth behavior behavior would be be expected ultimately to result in detection of of leakage prior to the possibility possibility of unstable pipe rupture. The periodic visual examinations for evidence ofof leakage that are performed during every refueling outage for the hot leg drain nozzle per per ASME Code Code Case Case N-722-1 are direct examinations of the metal surface that are capable of of detecting detecting small small amounts of of pressure pressure boundary leakage.
- Finally, Finally, the potential potential presence presence of weld weld repairs made during plant construction would not affect these conclusions. Any such such weld weld repairs repairs would have been made made prior to PHWT being applied, applied, and would be expected be expected to to extend extend over over aa relatively relatively limited limited circumferential portion of of the the original original weld.
weld. The The PWHT PWHT would would relax relax the the residual residual stresses stresses in the weld repair repair area, area, including including thethe substantial substantial relaxation expected expected atat the the surface surface exposed exposed to to primary primary coolant. coolant. Moreover, Moreover, in in the the unlikely unlikely case that initiation that initiation occurred occurred inin the the area of aa weld area of weld repair, repair, the weld weld repair repair would would bebe an an additional additional source source of of non-axisymmetric non-axisymmetric crack crack loading loading that that would tend to would tend to drive drive crack crack growth growth e StructuraI Structural Integrity Associates, 1nc Integrity Associates, Inc."
Mr. William Mr. William Sims Sims February February 25,2014 25, 2014 RAM-14-008 RAM-14-008 Page Page 55 of 16 ofl6 through-wall over through-wall over aa relatively local local circumferential circumferential region, region, ultimately ultimately resulting resulting inin detection detection of of leakage leakage prior to the possibility the possibility ofof unstable rupture. pipe rupture. These conclusions These conclusions extend extend to the the pipe pipe connection connection Alloy Alloy 821182 82/182 full full penetration penetration weldments weldments on on the the reactor cold reactor cold legs at Palisades. The assessments at Palisades. The assessments presented above above for the single for the single hot hot leg leg location clearly bound the concern for PWSCC PWSCC at at each each of of the cold leg locations. locations. The The susceptibility to PWSCC initiation is greatly reduced for nickel-based weldments weidments operating operating at reactor cold cold leg leg temperature, andand the PWSCC growth rate at at reactor cold leg temperature is approximately four times lower than the corresponding crack growth rate at reactor hot leg temperature (considering (considering thermal activation the standard thennal activation energy for crack growth of 130 130 kllmole kJ/mole per MRP-115). The PWHT applied to the cold leg locations is expected to result in similar similar residual stress levels as those calculated for the hot leg drain nozzle. lJ Structural Structural IntegrIty Associates, lnc Integrity Associates, Inc'"
Mr. William Mr. William Sims Sims February February 25,25, 2014 2014 RAM- 14-008 RAM-14-008 Page 6 of 16 Page 6 of 16 1 Figure Figure 1.1. Finite Finite Element Element Model Model e Slruclurallnlegrily Structural Associates, Inc Integrity Associates, Inc.-
Mr. William Mr. William Sims Sims February February 25,25, 2014 2014 RAM- 14-008 RAM-14-008 Page Page 77 ofof 16 16 Figure Figure 2.
- 2. Weld Weld Nuggets Nuggets e Structural Integrity Structural Integrity Associates, Inc.e Associates, Inc
Mr. William Sims Mr. William Sims February February 25,25, 2014 2014 RAM-14-008 RAM-14-008 Page 8 of16 Page 8 of 16 Figure 3. Patch Figure 3. Patch Nuggets Nuggets e Struclurallnlegrity Structural Associates, Inc Integrity Associates, Inc.-
Mr. William Mr. William Sims Sims February February 25,25, 2014 2014 RAM- 14-008 RAM-14-008 Page 9 Page 9 of 16 of 16 Figure Figure 4.
- 4. Radial Radial Stresses Stresses at at Operating Operating Conditions Conditions e Structurallntegrily StructuraI Associates, mc?
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Mr. William Mr. William Sims Sims February February 25,2014 25, 2014 RAM- 14-008 RAM-I 4-008 Page Page 1515 ofl6 of 16 1 ANSYS ANSYS~14.S EI.EM!NI'S R14.5 FEB FEB 24 24 2014 2014 11 : 14 : 39 11:14:39 PLF ro. PLOr NO. 11 Limit Analvsi.s Limit Analvsis Figure Figure 10.
- 10. Extent of Partial-Arc Extent of Partial-Arc Through-wal Through-walll Circumferen tial Flaw Circumferential Flaw Assumed Assumed for Limit Limit Analysis Analysis (ASME (ASME Code, Code, Level Level AA Service Service Limits)
Limits) Note: Note: Flaw Flaw isis Located Located at at the the Nozzle-Weld Nozzle-Weld Interface. Interface. lJ StructuralIntegrity fStructuraI Integrity Associates, Associates, Inc Inc.-
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