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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)
	v • d • e

{{#Wiki_filter:Entergy Nuclear Operations, Inc. Palisades Nuclear Plant - liritei 27780 Blue Star Memorial Highway Covert, Ml 49043-9530 Tel 269 764 2000 Anthony J. Vitale Site Vice President PNP 2014-015 February 25, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Relief Request Number RR 4-18 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 ISI 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 CFR 50.55a(g)(6)(ii)(F)(1) and 10 CFR 50.55a(g)(6)(ii)(F)(3), dated June 21, 2011. To support the startup of PNP following the current refueling outage, ENO requests approval of this alternative by March 8, 2014. This submittal contains no proprietary information. Summary of Commitments This letter identifies one new commitment, as described in Attachment 2, and no revised commitments. PNP 2014-015 February 25, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Entergy Nuclear Operations, Inc. Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530 Tel 269 764 2000 Anthony J. Vitale Site Vice President

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 lSI 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 CFR 50.55a(g)(6)(ii)(F)(1) and 10 CFR 50.55a(g)(6)(ii)(F)(3), dated June 21, 2011. To support the startup of PNP following the current refueling outage, ENO requests approval of this alternative by March 8, 2014. This submittal contains no proprietary information. Summary of Commitments This letter identifies one new commitment, as described in Attachment 2, and no revised commitments.

PNP 201 4-015 Page 2 Sincerely, /k44V4 ajv/jse Attachments: 1. Relief Request Number RR 4-18 Proposed Alternative

2. Description of Commitment

3. Structural Integrity Associates, Inc. Memorandum cc:

Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC PNP 2014-015 Page 2 Sincerely, ajv/jse Attachments:

1. Relief Request Number RR 4-18 Proposed Alternative

2. Description of Commitment

3. Structural Integrity Associates, Inc. Memorandum cc:

Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC. PALISADES NUCLEAR PLANT RELIEF REQUEST NUMBER RR 4-18 PROPOSED ALTERNATIVE in Accordance with 10 CFR 50.55a(a)(3)(ii) Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality and Safety 1. ASME Code Component(s Affected I Applicable Code Edition Components / Numbers: See Enclosure Table 1 Pressure Retaining Dissimilar Metal Piping Butt Welds Containing Alloy 82/182 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 Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W861 82 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section Xl, Division 1 N-770-1 Inspection Item: A-2 and B

== Description:== Class 1 Pressurized Water Reactor (PWR) pressure retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds containing Alloy 82/182 Unit / Inspection Interval: Palisades Nuclear Plant (PNP) / Fourth 10-Year Interval December 13, 2006 through December 12, 2015

===2. Applicable Code Requirements=== The ASME Boiler and Pressure Vessel Code, Rules for lnservice Inspection of Nuclear Power Plant Components, Section Xl, 2001 Edition through 2003 Addenda, as amended by 10 CFR 50.55a. With the issuance of a revised 10 CFR 50.55a in June 2011, the Nuclear Regulatory Commission (NRC) staff incorporated, by reference, Code Case N-770-1. Specific implementing requirements are documented in 10 CFR 50.55a(g)(6)(ii)(F) and are listed below: A. Regulation 10 CFR 50.55a(g)(6)(ii)(F)(1) states Licensees of existing, operating pressurized water reactors as of July 21, 2011 shall implement the requirements of ASME Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) of this section, by the first refueling outage after August 22, 2011. 1 of 8 ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC. PALISADES NUCLEAR PLANT RELIEF REQUEST NUMBER RR 4-18 PROPOSED ALTERNATIVE in Accordance with 10 CFR 50.55a(a)(3)(ii) Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality and Safety

1. ASME Code Component(s) Affected I Applicable Code Edition Components / Numbers:

See Enclosure Table 1 Pressure Retaining Dissimilar Metal Piping Butt Welds 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 Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1" N-770-1 Inspection Item: A-2 and B

== Description:== Class 1 Pressurized Water Reactor (PWR) pressure retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds containing Alloy 821182 Unit / Inspection Interval: Palisades Nuclear Plant (PNP) / Fourth 10-Year Interval December 13, 2006 through December 12, 2015

2. Applicable Code Requirements

The ASME Boiler and Pressure Vessel Code, Rules for Inservice Inspection of Nuclear Power Plant Components, Section XI, 2001 Edition through 2003 Addenda, as amended by 10 CFR 50.55a. With the issuance of a revised 10 CFR 50.55a in June 2011, the Nuclear Regulatory Commission (NRC) staff incorporated, by reference, Code Case N-770-1. Specific implementing requirements are documented in 10 CFR 50.55a(g)(6)(ii)(F) and are listed below: A. Regulation 10 CFR 50.55a(g)(6)(ii)(F)(1) states "Licensees of existing, operating pressurized water reactors as of July 21, 2011 shall implement the requirements of ASME Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) of this section, by the first refueling outage after August 22, 2011." 1 of 8

PROPOSED ALTERNATIVE B. Regulation 10 CFR 50.55a(g)(6)(ii)(F)(3) states that baseline examinations for welds in Code Case N-770-1, Table 1, Inspection Items A-i, A-2, and B, shall be completed by the end of the next refueling outage after January 20, 2012. The welds covered by this proposed alternative would be classified as Inspection Items A-2 and B (described below) for which visual and essentially 100 percent volumetric examination, as amended by 10 CFR 50.55a(g)(6)(ii)(F)(4), in part, are required. ASME Code Case N-770-1, Table 1, Examination Categories, as amended by 10 CFR 50.55a(g)(6)(ii)(F) CLASS 1 PWR Pressure Retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds Containing Alloy 82/182 Insp Extent and Frequency of Examination Parts Examined Item Bare metal visual examination each refueling outage. Unmitigated butt weld Essentially 100% volumetric examination for axial and at Hot Leg operating A-2 circumferential flaws in accordance with the applicable temperature (-2410) requirements of ASME Section XI, Appendix VIII, every five 625°F (329°C) years. Baseline examinations shall be completed by the end of the next refueling outage after January 20, 2012. Bare metal visual examination once per interval. Unmitigated butt weld Essentially 100% volumetric examination for axial and at Cold Leg operating circumferential flaws in accordance with the applicable temperature (-2410) B requirements of ASME Section XI, Appendix VIII, every second 525°F (274°C) and < inspection period not to exceed 7 years. Baseline examinations 580°F (304°C) shall be completed by the end of the next refueling outage after January 20, 2012. ASME Section Xl, Appendix VIII, Supplement 10, Qualification Requirements for Dissimilar Metal Piping Welds, is applicable to dissimilar metal (DM) welds without cast materials.

===3. Reason for Request=== Examinations of the DM welds listed in Enclosure Table 1 of this request could not be perlormed as required by ASME Code Case N-770-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(F). These DM welds are nominal pipe size (NPS) 2 inches and greater full penetration branch connection welds installed in primary coolant loop piping. See the Enclosure for typical configuration. The relevant conditions for this request for alternative are ASME Section Xl Code Case N-770-1, and 10 CFR 50.55a(g)(6)(ii)(F) items (1) and (3), which address performing the required baseline examinations. 2 of 8 PROPOSED ALTERNATIVE

8. Regulation 10 CFR 50.55a(g)(6)(ii)(F)(3) states that baseline examinations for welds in Code Case N-770-1, Table 1, Inspection Items A-1, A-2, and 8, shall be completed by the end of the next refueling outage after January 20, 2012.

The welds covered by this proposed alternative would be classified as Inspection Items A-2 and 8 (described below) for which visual and essentially 100 percent volumetric examination, as amended by 10 CFR 50.55a(g)(6)(ii)(F)(4), in part, are required. ASME Code Case N-77G-1, Table 1, Examination Categories, a8 amended by 10 CFR 50.558(g)(6)(II)(F) CLASS 1 PWR Pressure Retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds Containing Alloy 821182 Parts Examined Insp Extent and Frequency of Examination Item Bare metal visual examination each refueling outage. Unmitigated butt weld Essentially 100% volumetric examination for axial and at Hot Leg operating A-2 circumferential flaws in accordance with the applicable temperature (-2410) S requirements of ASME Section XI, Appendix VIII, every five 625°F (329°C) years. Baseline examinations shall be completed by the end of the next refueling outage after January 20,2012. Bare metal visual examination once per interval. Unmitigated butt weld Essentially 100% volumetric examination for axial and at Cold Leg operating circumferential flaws in accordance with the applicable temperature (-2410) 2! B requirements of ASME Section XI, Appendix VIII, every second 525°F (274°C) and < inspection period not to exceed 7 years. Baseline examinations 580°F (304°C) shall be completed by the end of the next refueling outage after January 20, 2012. ASME Section XI, Appendix VIII, Supplement 10, "Qualification Requirements for Dissimilar Metal Piping Welds," is applicable to dissimilar metal (OM) welds without cast materials.

3. Reason for Request

Examinations of the OM welds listed in Enclosure Table 1 of this request could not be performed as required by ASME Code Case N-770-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(F). These OM welds are nominal pipe size (NPS) 2 inches and greater full penetration branch connection welds installed in primary coolant loop piping. See the Enclosure for typical configuration. The relevant conditions for this request for alternative are ASME Section XI Code Case N-770-1, and 10 CFR 50.55a(g)(6)(ii)(F) items (1) and (3), which address performing the required baseline examinations. 2 of 8 ~

PROPOSED ALTERNATIVE Regulation 10 CFR 50.55a(g)(6)(ii)(F)(1) requires that licensees implement the requirements of ASME Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(1 0) of this section, by the first refueling outage after August 22, 2011. Regulation 10 CFR 50.55a(g)(6)(ii)(F)(3) requires that baseline examinations for welds in Code Case N-770-1 Table 1, Inspection Items A-i, A-2, and B be completed by the end of the next refueling outage after January 20, 2012. Relief is requested from 10 CFR 50.55a(g)(6)(ii)(F) items (1) and (3) for performance of required baseline volumetric examinations of the eight cold 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 (PDI) demonstrated joint configurations. That is, there currently are no PDI demonstrated volumetric techniques for this PNP weld joint configuration. Volumetric examination techniques for this complex weld configuration (see Enclosure Figures 1 and 2) require development and demonstration through the PDI qualification program. Under the PDI program, representative mock-ups would need to be fabricated in accordance with the 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 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 strategy has not been approved for this weld configuration. Creating a mitigation strategy would require significant time for development of tooling, qualification of procedures and personnel, and mockup verification, in order to develop an approved method that would also minimize radiation exposure to personnel. Due to the location of the welds, performing manual phased array ultrasonic test examinations of the nine welds would involve significant radiation exposure to personnel. Total dose incurred by examination, radiation protection, and supervisory personnel during ultrasonic testing of the nine weld locations is estimated to be at least 37 rem. This total includes preparation activities, and credits dose reduction controls and measures such as shielding, decontamination of components, high efficiency particulate air filter ventilation units, cameras, and remote telemetry. Dose rates vary, on contact, from 120 mremlhour to 15000 mrem/hour for the weld locations. General area dose rates in the vicinity of the weld locations vary from 25 mrem/hour to 200 mrem/hour. Development of a qualified procedure for the volumetric inspections would include 3 of 8 PROPOSED ALTERNATIVE Regulation 10 CFR 50.55a(g)(6)(ii)(F)(1) requires that licensees implement the requirements of ASME Code Case N-nO-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10 ) of this section, by the first refueling outage after August 22, 2011. Regulation 10 CFR 50.55a(g)(6)(ii)(F)(3) requires that baseline examinations for welds in Code Case N-770-1 Table 1, Inspection Items A-1, A-2, and 8 be completed by the end of the next refueling outage after January 20,2012. Relief is requested from 10 CFR 50.55a(g)(6)(ii)(F) items (1) and (3) for performance of required baseline volumetric examinations of the eight cold 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 (POI) demonstrated joint configurations. That is, there currently are no POI demonstrated volumetric techniques for this PNP weld joint configuration. Volumetric examination techniques for this complex weld configuration (see Enclosure Figures 1 and 2) require development and demonstration through the POI qualification program. Under the POI program, representative mock-ups would need to be fabricated in accordance with the POI 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 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 strategy has not been approved for this weld configuration. Creating a mitigation strategy would require significant time for development of tooling, qualification of procedures and personnel, and mockup verification, in order to develop an approved method that would also minimize radiation exposure to personnel. Due to the location of the welds, performing manual phased array ultrasonic test examinations of the nine welds would involve significant radiation exposure to personnel. Total dose incurred by examination, radiation protection, and supervisory personnel during ultrasonic testing of the nine weld locations is estimated to be at least 37 rem. This total includes preparation activities, and credits dose reduction controls and measures such as shielding, decontamination of components, high efficiency particulate air filter ventilation units, cameras, and remote telemetry. Dose rates vary, on contact, from 120 mremlhour to 15000 mrem/hour for the weld locations. General area dose rates in the vicinity of the weld locations vary from 25 mrernlhour to 200 mrem/hour. Development of a qualified procedure for the volumetric inspections would include 3 of 8

PROPOSED ALTERNATIVE identification of examination techniques that would minimize radiological exposure to examination personnel.

===4. Proposed Alternative and Basis for Use=== Proposed Alternative 1) Perform periodic system leakage tests in accordance with ASME Section Xl Examination Category B-P, Table IWB-2500-1 (Reference 10). 2) Perform visual and dye penetrant surface examinations of the welds in accordance with ASME requirements. During the 2012 (1R22) and 2014 (1R23) refueling outages, visual and external surface examinations of certain welds for which relief is requested identified no evidence of through-wall cracking or leakage for these components, as identified in Enclosure Table 1. Pursuant to 10 CFR 50.55a(3)(ii), ENO proposes to perform appropriate actions to meet ASME Section Xl Code Case N-770-1 examination requirements, as required, for those dissimilar metal welds identified in Enclosure Table 1 of this request during the first refueling outage after a viable technology is developed to perform these examinations. Basis for Use Enclosure Table 1 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, as required by 10 CFR 50.55a(g)(6)(ii)(F) items (1) and (3). Examination History Enclosure Table 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, for the three weld locations that were not subject to surface or visual examinations during the ongoing 1 R23 refueling outage (i.e., weld no.s 3, 6, and 7 in Enclosure Table 1), maintenance activities in the vicinity of the weld locations during the 1 R23 refueling outage did not identify observations of leakage from the welds. Structural Evaluation Structural Integrity Associates, Inc. (SIA) performed a bounding evaluation of the Alloy 82/1 82 full penetration weld which connects the hot leg to the drain nozzle (see SIA. Memorandum in Attachment 3). The evaluation concluded the following: Because of the post weld heat treatment (PWHT) of the nickel-base materials, there is a low probability that a PWSCC crack of engineering size has initiated on the Alloy 82/182 full-penetration branch pipe connection welds at PNP. Engineering size is defined as a flaw with a depth of about 1 to 2 millimeters. 4 of 8 PROPOSED ALTERNATIVE identification of examination techniques that would minimize radiological exposure to examination personnel.

4. Proposed Alternative and Basis for Use

Proposed Alternative

1) Perform periodic system leakage tests in accordance with ASME Section XI Examination Category B-P, Table IWB-2S00-1 (Reference 10).

2) Perform visual and dye penetrant surface examinations of the welds in accordance with ASME requirements. During the 2012 (1R22) and 2014 (1R23) refueling outages, visual and external surface examinations of certain welds for which relief is requested identified no evidence of through-wall cracking or leakage for these components, as identified in Enclosure Table 1.

Pursuant to 10 CFR SO.SSa(3)(ii), ENO proposes to perform appropriate actions to meet ASME Section XI Code Case N-770-1 examination requirements, as required, for those dissimilar metal welds identified in Enclosure Table 1 of this request during the first refueling outage after a viable technology is developed to perform these examinations. Basis for Use Enclosure Table 1 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, as required by 10 CFR SO.SSa(g)(6)(ii)(F) items (1) and (3). Examination History Enclosure Table 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, for the three weld locations that were not subject to surface or visual examinations during the ongoing 1 R23 refueling outage (Le., weld no.'s 3, 6, and 7 in Enclosure Table 1), maintenance activities in the vicinity of the weld locations during the 1 R23 refueling outage did not identify observations of leakage from the welds. Structural Evaluation Structural Integrity Associates, Inc. (SIA) performed a bounding evaluation of the Alloy 821182 full penetration weld which connects the hot leg to the drain nozzle (see SIA Memorandum in Attachment 3). The evaluation concluded the following: Because of the post weld heat treatment (PWHT) of the nickel-base materials, there is a low probability that a PWSCC crack of engineering size has initiated on the Alloy 821182 full-penetration branch pipe connection welds at PNP. Engineering size is defined as a flaw with a depth of about 1 to 2 millimeters. 4 of 8

PROPOSED ALTERNATIVE The finite-element calculations for the hot-leg drain nozzle show relatively modest peak total tensile stresses on the wetted surface for normal operating conditions due to the benefit of the PWHT applied. ASME Code acceptance criteria are satisfied for 60 effective full power years for a circumferential flaw, and more than 34 years for an axial flaw assuming crack initiates at day one. Using hot leg crack growth rate and temperature. o PWHT reduced the crack growth rate for Alloy 182 weld metal (e.g., by a factor of between two and four). This benefit is conservatively not credited in the MRP-1 15 crack growth rate equation for Alloy 182. o The susceptibility to PWSCC initiation is greatly reduced for nickel-based weldments operating at reactor 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 temperature An additional limit analysis was performed for the hypothetical partial-arc through-wall circumferential flaw illustrated in Figure 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 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 this conservatively assumed axial flaw is shown in Figure 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: used two times upset pressure 2650 psi.) Operating Conditions The operating temperature of a component is a primary factor influencing the initiation of PWSCC. Research by the Electric Power Research Institute (EPRI) (Reference 8) indicates that the difference in the operating temperature between hot leg locations and cold leg locations is sufficient to significantly influence the time to initiation of PWSCC, with the susceptibility increasing with temperature. The research reports PWSCC is least likely to occur in cold leg temperature penetrations. All but one of the welds covered by this relief are primarily found in lower temperature regions of the system, typically at temperatures near to Tcold, which is approximately 537 °F. This means, for these welds, there is a lower probability of crack initiation, and a slower crack growth rate (Reference 9). 5 of 8 PROPOSED ALTERNATIVE The finite-element calculations for the hot-leg drain nozzle show relatively modest peak total tensile stresses on the wetted surface for normal operating conditions due to the benefit of the PWHT applied. ASME Code acceptance criteria are satisfied for 60 effective full power years for a circumferential flaw, and more than 34 years for an axial flaw assuming crack initiates at day one. Using hot leg crack growth rate and temperature. o PWHT reduced the crack growth rate for Alloy 182 weld metal (e.g., by a factor of between two and four). This benefit is conservatively not credited in the MRP-115 crack growth rate equation for Alloy 182. o The susceptibility to PWSCC initiation is greatly reduced for nickel-based weldments operating at reactor 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 temperature An additional limit analysis was performed for the hypothetical partial-arc through-wall circumferential flaw illustrated in Figure 10 in Attachment 3. This flaw is predicted to be through-wall for 45 degrees in approximately 100 years using the MRP-115 crack growth rate equation. In the unlikely case of initiation of a 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 821182 weld cross section and a large portion of the Alloy 600 nozzle. The extent of this conservatively assumed axial flaw is shown in Figure 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: used two times upset pressure 2650 psi.) Operating Conditions The operating temperature of a component is a primary factor influencing the initiation of PWSCC. Research by the Electric Power Research Institute (EPRI) (Reference 8) indicates that the difference in the operating temperature between hot leg locations and cold leg locations is sufficient to significantly influence the time to initiation of PWSCC, with the susceptibility increasing with temperature. The research reports PWSCC is least likely to occur in cold leg temperature penetrations. All but one of the welds covered by this relief are primarily found in lower temperature regions of the system, typically at temperatures near to Tcold, which is approximately 537 of. This means, for these welds, there is a lower probability of crack initiation, and a slower crack growth rate (Reference 9). 5 of 8

PROPOSED ALTERNATIVE Leakage Detection Capabilities Even if there were to be flaws in the welds for which relief is requested, and these flaws led to leakage, the leak detection methodology presently used by industry is very sensitive. After a number of recent operating events, the industry imposed an NEI 03-08, Guideline for the Management of Materials Issues, requirement to improve leak detection capability. As a result, virtually all pressurized water reactors (PWRs) in the United States, including PNP, have a leak detection capability of less than or equal to 0.1 gpm (Reference 7). All plants, including PNP, also monitor seven-day moving averages of reactor coolant system leak rates. Action response times following a detected primary coolant system leak vary, 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) report, WCAP-1 6465, and are stated below: 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 - One seven day rolling average of unidentified RCS inventory balance values greater than 0.1 gpm. Level 2 - Two consecutive unidentified RCS inventory balance values greater than 0.15 gpm. Level 3 - One unidentified RCS inventory balance value greater than 0.3 gpm. Note: Calculation of the absolute RCS inventory balance values must include the rules for the treatment of negative values and missing observations. Action levels on the deviation from the baseline mean: Level 1 - Nine consecutive unidentified RCS inventory balance values greater than the baseline mean [pJ value. Level 2 - Two of three consecutive unidentified RCS inventory balance values greater than [p + 2o], where a is the baseline standard deviation. Level 3 - One unidentified RCS inventory balance value greater than [p +3o]. These action levels have been incorporated into PNP procedures. 6 of 8 PROPOSED ALTERNATIVE Leakage Detection Capabilities Even if there were to be flaws in the welds for which relief is requested, and these flaws led to leakage, the leak detection methodology presently used by industry is very sensitive. After a number of recent operating events, the industry imposed an NEI 03-08, "Guideline for the Management of Materials Issues," requirement to improve leak detection capability. As a result, virtually all pressurized water reactors (PWRs) in the United States, including PNP, have a leak detection capability of less than or equal to 0.1 gpm (Reference 7). All plants, including PNP, also monitor seven-day moving averages of reactor coolant system leak rates. Action response times following a detected primary coolant system leak vary, 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) report, WCAP-16465, and are stated below: 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 - One seven day rolling average of unidentified RCS inventory balance values greater than 0.1 gpm. Level 2 - Two consecutive unidentified RCS inventory balance values greater than 0.15 gpm. Level 3 - One unidentified RCS inventory balance value greater than 0.3 gpm. Note: Calculation of the absolute RCS inventory balance values must include the rules for the treatment of negative values and missing observations. Action levels on the deviation from the baseline mean: Level 1 - Nine consecutive unidentified RCS inventory balance values greater than the baseline mean [J,J] value. Level 2 - Two of three consecutive unidentified RCS inventory balance values greater than [J,J + 20], where 0 is the baseline standard deviation. Level 3 - One unidentified RCS inventory balance value greater than [J,J +30]. These action levels have been incorporated into PNP procedures. 6 of 8

PROPOSED ALTERNATIVE A small steam leak from a weld flaw would, over time, result in a rise in containment sump level rate of increase. Containment sump level is continually monitored, and if a rise in the rate of containment sump level increase is observed, plant procedures direct plant operators to identify the source of the leakage. Operators may also be alerted to a leak from a flaw by containment radiation monitoring instrumentation. This instrumentation, required by the Technical Specifications, is capable of detecting a 100 cm 3 /min leak in 45 minutes, based on 1% failed fuel. Periodic system leakage tests are performed in accordance with ASME Section Xl. Operator walkdowns of containment are periodically performed during power operations at lower levels of containment to detect leakage. Therefore, with the periodic system leakage tests, the visual and surface examinations performed during 2012 and 2014 refueling outages, the 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 and standards, July 25, 2013. 2. ASME Section XI, Rules For Inservice Inspection of Nuclear Power Plant Components, 2001 Edition with Addenda through 2003. 3. ASME Section Xl, Division 1, Code Case N-460, Alternative Examination Coverage for Class 1 and Class 2 Welds, Section Xl, Division 1. 4. Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-1 39), Revision 1, EPRI, Palo Alto, CA, 2008 [ADAMS Accession Number ML1009700671]. 5. Nondestructive Evaluation: Procedure for Manual Phased Array Ultrasonic Examination of Dissimilar Metal Welds, EPRI-DMW-PA-1, Revision 3, 1016645, EPRI Palo Alto, CA, 2008. 6. Changing the Frequency of Inspections for PWSCC Susceptible Welds at Cold Leg Temperatures, in Proceedings of 2011 ASME Pressure Vessels and Piping Conference, July 17-21, 2011, Baltimore, MD. 7. WCAP-16465-NP, Rev. 0, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors, Westinghouse Electric Co., September 2006 [ML070310082]. 8. Electric Power Research Institute: PWSCC of Alloy 600 Materials in PWR Primary System Penetrations, EPRI, Palo Alto, CA, 1994, TR-1 03696 [MLO131 10446]. 9. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-1 15), EPRI, Palo Alto, CA, 2004, 1006696 [MLO51 100204]. 7 of 8 PROPOSED ALTERNATIVE A small steam leak from a weld flaw would, over time, result in a rise in containment sump level rate of increase. Containment sump level is continually monitored, and if a rise in the rate of containment sump level increase is observed, plant procedures direct plant operators to identify the source of the leakage. Operators may also be alerted to a leak from a flaw by containment radiation monitoring instrumentation. This instrumentation, required by the Technical Specifications, is capable of detecting a 100 cm3/min leak in 45 minutes, based on 1 % failed fuel. Periodic system leakage tests are performed in accordance with ASME Section XI. Operator walkdowns of containment are periodically performed during power operations at lower levels of containment to detect leakage. Therefore, with the periodic system leakage tests, the visual and surface examinations performed during 2012 and 2014 refueling outages, the 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 and standards," July 25, 2013.

2. ASME Section XI, "Rules For Inservice Inspection of Nuclear Power Plant Components,"

2001 Edition with Addenda through 2003.

3. ASME Section XI, Division 1, Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 Welds, Section XI, Division 1."

4. Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139), Revision 1, EPRI, Palo Alto, CA, 2008 [ADAMS Accession Number ML1009700671].

5. Nondestructive Evaluation: Procedure for Manual Phased Array Ultrasonic Examination of Dissimilar Metal Welds, EPRI-DMW-PA-1, Revision 3,1016645, EPRI Palo Alto, CA, 2008.

6. "Changing the Frequency of Inspections for PWSCC Susceptible Welds at Cold Leg Temperatures", in Proceedings of 2011 ASME Pressure Vessels and Piping Conference, July 17-21, 2011, Baltimore, MD.

7. WCAP-16465-NP, Rev. 0, "Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors,"

Westinghouse Electric Co., September 2006 [ML070310082].

8. Electric Power Research Institute: PWSCC of Alloy 600 Materials in PWR Primary System Penetrations, EPRI, Palo Alto, CA, 1994, TR-103696 [ML013110446].

9. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82,182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA, 2004, 1006696 [ML051100204].

7 of 8

PROPOSED ALTERNATIVE

10. PNP Technical Specification Surveillance Procedure RT-71A, Primary Coolant System, Class 1 System Leakage Test, Revision 18.

11. Pressurized Water Reactor (PWR) Owners Group Letter OG-1 2-89, Transmittal of Final Relief Request Famework under Relief Request for Large Diameter Cold Leg Locations with Obstructions (PA-MSC-0934), March 8, 2012.

7. Enclosure Table 1 Weld Examination History Figure 1 - Nozzle Assembly Materials Figure 2 - Hot Leg Drain Nozzle Configuration (Representative) 8 of 8 PROPOSED ALTERNATIVE

10. PNP Technical Specification Surveillance Procedure RT-71A, "Primary Coolant System, Class 1 System Leakage Test," Revision 18.

11. Pressurized Water Reactor (PWR) Owner's Group Letter OG-12-89, "Transmittal of

'Final Relief Request Famework' under Relief Request for Large Diameter Cold Leg Locations with Obstructions (PA-MSC-0934)," March 8, 2012.

7. Enclosure Table 1 - Weld Examination History Figure 1 - Nozzle Assembly Materials Figure 2 - Hot Leg Drain Nozzle Configuration (Representative) 8 of 8

PROPOSED ALTERNATIVE ENCLOSURE Table 1 Weld Examination History No. Description ISI Weld ID Location 1R19 1R20 1R21 1R22 1R23 Examinations Examinations Examinations Examinations Examinations Visual Surface 2 inch Cold Leg PCS-30-RCL-1 A-P-50A Discharge (Report# 4046 (Report# 1 R23-1. Charging Nozzle 11/2 Leg Exam number PT-14-025) 06-26) Visual Surface 2 inch Cold Leg PCS-30-RCL-1A-P-50A Suction (Report# 4047 (Report# 1 R23-2. Drain Nozzle 5/2 Leg Exam number PT-i4-031) 07-28.1) Visual Surface 3 inch Cold Leg PCS-30-RCL-1 B-P-50B Discharge (Report# VT-b-(Report# 1 R22-3. Pressurizer Spray 10/3 Leg 069) PT-12-039) Nozzle Visual Surface 2 inch Cold Leg PCS-30-RCL-i B-P-50B Suction (Report# VT-i 0-(Report #1 R23-Drain Nozzle 5/2 Leg 048) PT-14-032) Visual Surface 2 inch Cold Leg PCS-30-RCL-2A-P-50C Discharge (Report# VT (Report# 1 R23-Charging Nozzle 11/2 Leg 083) PT-14-019) Visual Surface 3 inch Cold Leg PCS-30-RCL-2A-P-50C Discharge (Report# VT (Report# 1 R22-6. Pressurizer Spray 11/3 Leg Nozzle 035) PT-12-032) Visual 2 inch Cold Leg PCS-30-RCL-2A-P-50C Suction (Report# VT Drain Nozzle 5/2 Leg 038) Visual Surface 2 inch Cold Leg PCS-30-RCL-2B-P-50D Suction (Report# VT-i 0-(Report# 1 R23-8. Drain and 5/2 Leg 071) PT-i4-020) Letdown Nozzle Visual Visual Visual Visual Visual (Report# 1 R23-2 inch Hot Leg PCS-42-RCL-i H-A Hot Leg (Report# 4047 (Report# VT (Report# VT-i 0-(Report# 1 R22-VT-i 4-059) Drain Nozzle 3/2 Exam number 062) 022) VT-12-076) 07-23.1) PROPOSED ALTERNATIVE ENCLOSURE Table 1 Weld Examination History No. Description ISIWeldlD Location 1R19 1R20 1R21 1R22 1R23 Examinations Examinations Examinations Examinations Examinations Visual Surface

1.

2 inch Cold Leg PCS-30-RCL-l A-P-50A Discharge (Report# 4046 (Report# 1 R23-Charging Nozzle 11/2 Leg Exam number PT-14-025) 06-26) Visual Surface 2 inch Cold Leg PCS-30-RCL-l A-P-50A Suction (Report# 4047 (Report# 1 R23-

2.

Drain Nozzle 5/2 Leg Exam number PT-14-031) 07-28.1) 3 inch Cold Leg Visual Surface PCS-30-RCL-l B-P-50B Discharge (Report# VT (Report# 1 R22-

3.

Pressurizer Spray 10/3 Leg 069) PT-12-039) Nozzle Visual Surface 2 inch Cold Leg PCS-30-RCL -1 B-P-50B Suction (Report# VT-l0-(Report #1 R23-

4.

Drain Nozzle 5/2 Leg 048) PT-14-032) Visual Surface 2 inch Cold Leg PCS-30-RCL-2A-P-50C Discharge (Report# VT (Report# 1 R23-

5.

Charging Nozzle 11/2 Leg 083') PT-14-019) 3 inch Cold Leg Visual Surface PCS-30-RCL-2A-P-50C Discharge (Report# VT (Report# 1 R22-

6.

Pressurizer Spray 11/3 Leg 035) PT-12-032) Nozzle P-50C Suction Visual 2 inch Cold Leg PCS-30-RCL-2A-(Report# VT 7. Drain Nozzle 5/2 Leg 038) 2 inch Cold Leg Visual Surface PCS-30-RCL-2B-P-50D Suction (Report# VT-l0-(Report# 1 R23-

8.

Drain and 5/2 Leg 071) PT-14-020) Letdown Nozzle Visual Visual Visual Visual Visual (Report# 1 R23-2 inch Hot Leg PCS-42-RCL -1 H-A Hot Leg (Report# 4047 (Report# VT (Report# VT-l0-(Report# 1 R22-VT-14-059)

9.

Drain Nozzle 3/2 Exam number 062) 022) VT-12-076) 07-23.1)

PROPOSED ALTERNATIVE ENCLOSURE Figure 1 Nozzle Assembly Materials Not to Scale Stainless Steel Cladding lCold Leg Full Penetration DM Alloy 82/182 Weld PROPOSED ALTERNATIVE ENCLOSURE Figure 1 Nozzle Assembly Materials Alloy 821182 Stainless Steel Cladding 4----i Carbon Steel Cold Leg Full Penetration DM Alloy 821182 Weld Alloy 600 Nozzle Not to Scale

a corau c pi f 3ID 7P FI7 4Y FJL.DIN RAWE 5CK/

ra ScAIO 4ir4L WL PROPOSED ALTERNATIVE ENCLOSURE Figure 2 Hot Leg Drain Nozzle Configuration (Representative) (excerpt from PNP vendor drawing VEN-M1-D Sheet 108, Revision 8) IN IIOZ XALE.6/2 PROPOSED ALTERNATIVE ENCLOSURE Figure 2 Hot Leg Drain Nozzle Configuration (Representative) (excerpt from PNP vendor drawing VEN-M1-D Sheet 108, Revision 8) I DQAIN NOZZ,£. $741£ 6'./1'

ATTACHMENT 2 DESCRIPTION OF COMMITMENT This table identifies actions discussed in this letter for which Entergy Nuclear Operations, Inc. (ENO) commits to perform. Any other actions discussed in this submittal are described for information only and 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 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. these examinations. ATTACHMENT 2 DESCRIPTION OF COMMITMENT This table identifies actions discussed in this letter for which Entergy Nuclear Operations, Inc. (ENO) commits to perform. Any other actions discussed in this submittal are described for information only and are not commitments. TYPE SCHEDULED (Check one) COMPLETION COMMITMENT ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required) ENO will perform appropriate actions to meet ./ The first ASME Section XI 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. these examinations.

ATTACHMENT 3 STRUCTURAL INTEGRITY ASSOCIATES, INC. MEMORANDUM Evaluation of the Palisades Nuclear Plant Hot Leg Drain Nozzle for Primary Water Stress Corrosion Cracking February 25, 2014 RAM-i 4-008 16 Pages Follow ATTACHMENT 3 STRUCTURAL INTEGRITY ASSOCIATES, INC. MEMORANDUM Evaluation of the Palisades Nuclear Plant Hot Leg Drain Nozzle for Primary Water Stress Corrosion Cracking February 25, 2014 RAM-14-008 16 Pages Follow

Structural Integrity Associates, Inc. 5215 HellyerAve. Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.structint.com rmattson@suctintcom MEMORANDUM February 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 1 depicts the finite element model, and Figures 2 and 3 depict the weld and patch nuggets, respectively. Analyses were performed for the following steps of construction: 1. Deposit cladding on hot leg inside surface (ID). 2. Install drain line nozzle/backing ring and deposit weld. 3. Remove backing ring and deposit ID patch. 4. Post weld heat treatment, including creep effects based upon experimental data. 5. Subject the configuration to a hydrostatic test. 6. Impose five cycles of shake down at normal operating temperature and pressure. Stress results normal to a circumferential crack are shown in Figure 4, and those normal to an axial crack are shown in Figure 5. ToIl-Free 877-474-7693 Akron.OH Aibaqu.rque,NM Ausfln,TX Chartol6,NC Chattanaaga,Th chicaoon. 330-889-9753 505-872-0123 512-533-9191 704-597-5554 423-553-1180 815-648-2519 Donvec, CO Mystic, CT PoughkeepsW NY San Dig., CA San Joan, CA Stat. Colige, PA Thtonto, Canada 303-792-0077 880.536-3982 845-454-6100 858.4556360 408-978-8280 814-954-7776 905629-98l7 S) Structural Integrity Associates, Inc. February 25,2014 RAM-14-008 TO: FROM: MEMORANDUM William Sims Dick Mattson 5215 Hellyer Ave. Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.structinlcom rmattson@structinlcom

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 1 depicts the finite element model, and Figures 2 and 3 depict the weld and patch nuggets, respectively. Analyses were performed for the following steps of construction:

1. Deposit cladding on hot leg inside surface (lD).

2. Install drain line nozzlelbacking ring and deposit weld.

3. Remove backing ring and deposit ID patch.

4. Post weld heat treatment, including creep effects based upon experimental data.

5. Subject the configuration to a hydrostatic test.

6. Impose five cycles of "shake down" at normal operating temperature and pressure.

Stress results normal to a circumferential crack are shown in Figure 4, and those normal to an axial crack are shown in Figure 5. _________________ Toll-Free 877-474-7693 _______________ _ AIIran, 011 33IJ.88&.9753 ~. AusDl,tx 505.an.Q123 512-533-9191 .... CT f'aaIIIIbIpsIt, NY -'536-3l1li2 845-454-6100 ChnIIII, III 704-597-5554 423-553*1180 81~2519 8tIIt CIIIIgI. PA __ CInadI 814-954-7776 9Q5.829.9817

Mr. William Sims February 25, 2014 RAM-14-008 Page 2 of 16 Crack tip elements along the nozzle-weld boundary were evaluated at seven depths through the thickness, and at 00, 30°, 60° and 90° angular locations, for a postulated circumferential flaw. A plot of the applied stress intensity vs. crack depth is shown in Figure 6 for the circumferential flaw. Assessment of Likelihood of PWSCC Initiation The likelihood of PWSCC initiation occurring on the wetted Alloy 600 and Alloy 82/1 82 surfaces of the Palisades hot leg drain nozzle was assessed by Dominion Engineering, Inc. (DEl). This assessment concludes that there is a low probability that a stress corrosion crack of engineering size has initiated in the Alloy 82/182 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 of PWSCC of Alloy 600 weldments that have been exposed to PWHT after welding is demonstrated by the following: detailed laboratory investigations including studies that show a very significant relaxation 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). 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 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 circumferential and axial flaws was investigated in order to assess the consequences of such hypothetical cracking in the unlikely case of PWSCC initiation. Using the applied stress intensity factors described above for a circumferential flaw, crack growth through the depth of the weld was calculated, and the results are presented in Figure 7 for the circumferential flaw. For the axial flaw, a conservative classical fracture mechanics solution was used for an elliptical flaw which has a constant width along the inside surface of the Alloy 600 nozzle and Alloy Letter from G. White (Dominion Engineering, Inc.) to W. Sims (Entergy), Effect of Post-Weld Heat Treatment Applied to Alloy 82/1 82 Full-Penetration Branch Pipe Connection Welds at Palisades, L-4199-OO-O1, Rev. 0, dated February 25, 2014. Structura! Integrity Associates, Inc. Mr. William Sims RAM-14-008 February 25,2014 Page 2 of 16 Crack tip elements along the nozzle-weld boundary were evaluated at seven depths through the thickness, and at 0°,30°, 60° and 90° angular locations, for a postulated circumferential flaw. A plot of the applied stress intensity vs. crack depth is shown in Figure 6 for the circumferential flaw. Assessment of Likelihood of PWSCC Initiation The likelihood of PWSCC initiation occurring on the wetted Alloy 600 and Alloy 821182 surfaces of the Palisades hot leg drain nozzle was assessed by Dominion Engineering, Inc. (DEI)). This assessment concludes that there is a low probability that a stress corrosion crack of engineering size has initiated in the Alloy 82/182 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 Alloy 600 weldments that have been exposed to PWHT after welding is demonstrated by the following: detailed laboratory investigations including studies that show a very significant relaxation 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). 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 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 circumferential and axial flaws was investigated in order to assess the consequences of such hypothetical cracking in the unlikely case of PWSCC initiation. Using the applied stress intensity factors described above for a circumferential flaw, crack growth through the depth of the weld was calculated, and the results are presented in Figure 7 for the circumferential flaw. For the axial flaw, a conservative classical fracture mechanics solution was used for an elliptical flaw which has a constant width along the inside surface of the Alloy 600 nozzle and Alloy I Letter from G. White (Dominion Engineering, Inc.) to W. Sims (Entergy), "Effect of Post-Weld Heat Treatment Applied to Alloy 82/182 Full-Penetration Branch Pipe Connection Welds at Palisades," L-4199-00-01, Rev. 0, dated February 25, 2014. l) Structural Integrity Associates, Inc.e

Mr. William Sims February 25, 2014 RAM-14-008 Page 3 of 16 82/182 weld, and a variable depth through the thickness. Three widths for the elliptical flaw were modeled. Figures 8 and 9 present the applied stress intensity factors vs. crack depth and crack growth for the axial flaw, respectively. However, more realistic, less conservative results can be obtained using finite element analysis techniques for the axial flaw as was done for the circumferential flaw. The crack growth law used in developing these plots is as described in MRP-1 15. As discussed in the DEl letter, laboratory crack growth testing has shown a significant benefit of PWHT in reducing the crack growth rate for Alloy 182 weld metal (e.g., by a factor of between two and four). This benefit is conservatively not credited in the MRP-1 15 crack growth rate equation for Alloy 182. Limit Analysis Circumferential Cracking: At 60 effective full power years, the finite element model described above was modified to include cracking at the four circumferential locations. A limit analysis, as described in ASME Code, Section III, Subparagraph NB-3228. 1, was performed, taking into consideration the effects of the flux Alloy 82/182 weld by decreasing the weld metal effective yield strength by the Z factor. The limit analysis results satisfy the Section III criteria. In the unlikely event that circumferential PWSCC were to occur in the Alloy 82/182 full 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 circumference. This non-axisymmetric crack growth behavior would be expected ultimately to result in detection of leakage prior to the possibility of unstable pipe rupture. An additional limit analysis was performed for the hypothetical partial-arc through-wall circumferential flaw illustrated in Figure 10. This flaw is predicted to be through-wall for 45° in approximately 100 years using the MRP-1 15 crack growth rate equation. The analysis, which applied Level A Service Limits of the ASME Code, showed that the flaw remains stable at twice the applied loads. Axial Cracking: For the axial flaw, the flaw modeled in ANSYS was conservatively assumed to be at a depth of 75% through-wall. The analysis satisfies the Section III criteria. Per Figure 9, this equates to 34 effective full power years of operation. As noted above, if more detailed finite element analyses were performed, there would be an increase in the effective full power years of operation corresponding to a flaw depth of 75% through-wall. 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 this conservatively assumed axial flaw is shown in Figure 11. The analysis, which applied Level A Service Limits of the ASME Code, showed that the flaw remains stable at 1.87 times the applied StructoraI Integrity Associates, Inc Mr. William Sims RAM-14-008 February 25,2014 Page 3 of 16 82/182 weld, and a variable depth through the thickness. Three widths for the elliptical flaw were modeled. Figures 8 and 9 present the applied stress intensity factors vs. crack depth and crack growth for the axial flaw, respectively. However, more realistic, less conservative results can be obtained using finite element analysis techniques for the axial flaw as was done for the circumferential flaw. The crack growth law used in developing these plots is as described in MRP-115. As discussed in the DEI letter, laboratory crack growth testing has shown a significant benefit of PWHT in reducing the crack growth rate for Alloy 182 weld metal (e.g., by a factor of between two and four). This benefit is conservatively not credited in the MRP-I 15 crack growth rate equation for Alloy 182. Limit Analysis Circumferential Cracking: At 60 effective full power years, the finite element model described above was modified to include "cracking" at the four circumferential locations. A limit analysis, as described in ASME Code, Section III, Subparagraph NB-3228.1, was performed, taking into consideration the effects of the flux Alloy 821182 weld by decreasing the weld metal effective yield strength by the "Z" factor. The limit analysis results satisfy the Section III criteria. In the unlikely event that circumferential PWSCC were to occur in the Alloy 82/182 full 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 circumference. This non-axisymmetric crack growth behavior would be expected ultimately to result in detection of leakage prior to the possibility of unstable pipe rupture. An additional limit analysis was performed for the hypothetical partial-arc through-wall circumferential flaw illustrated in Figure 10. This flaw is predicted to be through-wall for 45° in approximately 100 years using the MRP-l 15 crack growth rate equation. The analysis, which applied Level A Service Limits of the ASME Code, showed that the flaw remains stable at twice the applied loads. Axial Cracking: For the axial flaw, the flaw modeled in ANSYS was conservatively assumed to be at a depth of 75% through-wall. The analysis satisfies the Section III criteria. Per Figure 9, this equates to 34 effective full power years of operation. As noted above, if more detailed finite element analyses were performed, there would be an increase in the effective full power years of operation corresponding to a flaw depth of75% through-wall. 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 821182 weld cross-section and a large portion of the Alloy 600 nozzle. The extent of this conservatively assumed axial flaw is shown in Figure 1 I. The analysis, which applied Level A Service Limits of the ASME Code, showed that the flaw remains stable at 1.87 times the applied e Structural Integrity Associates, Inc'"

Mr. William Sims February 25, 2014 RAM-14-008 Page 4 of 16 loads. 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. Conclusions Based on the assessments and calculations performed, it is concluded that the visual examinations for evidence of pressure boundary leakage required by ASME Code Case N-722-1 are sufficient to ensure that nuclear safety is maintained (i.e., periodic volumetric/surface examinations for indications of PWSCC are not necessary). The adequacy 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 probability that a stress corrosion crack of engineering size has initiated on the Alloy 82/182 full 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 ofplant 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 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, in the unlikely case of initiation of a circumferential crack and the unlikely case that a circumferential crack were to exceed a depth of 60% through-wall, non-axisymmetric crack growth behavior would be expected ultimately to result in detection of leakage prior to the possibility of unstable pipe rupture. The periodic visual examinations for evidence of leakage that are performed during every refueling outage for the hot leg drain nozzle per ASME Code Case N-722-1 are direct examinations of the metal surface that are capable of detecting small amounts of pressure boundary leakage. Finally, the potential presence of weld repairs made during plant construction would not affect these conclusions. Any such weld repairs would have been made prior to PHWT being applied, and would be expected to extend over a relatively limited circumferential portion of the original weld. The PWHT would relax the residual stresses in the weld repair area, including the substantial relaxation expected at the surface exposed to primary coolant. Moreover, in the unlikely case that initiation occurred in the area of a weld repair, the weld repair would be an additional source of non-axisymmetric crack loading that would tend to drive crack growth StructuraI Integrity Associates, 1nc Mr. William Sims RAM-14-008 February 25,2014 Page 4 of 16 loads. 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. Conclusions Based on the assessments and calculations performed, it is concluded that the visual examinations for evidence of pressure boundary leakage required by ASME Code Case N-722-1 are sufficient to ensure that nuclear safety is maintained (i.e., periodic volumetric/surface examinations for indications of PWSCC are not necessary). The adequacy 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 probability that a stress corrosion crack of engineering size has initiated on the Alloy 82/182 full 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 of75% through-wall, 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. Similarly, in the unlikely case of initiation of a circumferential crack and the unlikely case that a circumferential crack were to exceed a depth of 60% through-wall, non-axisymmetric crack growth behavior would be expected ultimately to result in detection of leakage prior to the possibility of unstable pipe rupture. The periodic visual examinations for evidence of leakage that are performed during every refueling outage for the hot leg drain nozzle per ASME Code Case N-722-1 are direct examinations of the metal surface that are capable of detecting small amounts of pressure boundary leakage. Finally, the potential presence of weld repairs made during plant construction would not affect these conclusions. Any such weld repairs would have been made prior to PHWT being applied, and would be expected to extend over a relatively limited circumferential portion of the original weld. The PWHT would relax the residual stresses in the weld repair area, including the substantial relaxation expected at the surface exposed to primary coolant. Moreover, in the unlikely case that initiation occurred in the area of a weld repair, the weld repair would be an additional source of non-axisymmetric crack loading that would tend to drive crack growth e Structural Integrity Associates, Inc."

Mr. William Sims February 25, 2014 RAM-14-008 Page 5 ofl6 through-wall over a relatively local circumferential region, ultimately resulting in detection of leakage prior to the possibility of unstable pipe rupture. These conclusions extend to the pipe connection Alloy 82/182 full penetration weldments on the reactor cold legs at Palisades. The assessments presented above for the single hot leg location clearly bound the concern for PWSCC at each of the cold leg locations. The susceptibility to PWSCC initiation is greatly reduced for nickel-based weidments operating at reactor cold leg temperature, and the PWSCC growth rate at reactor cold leg temperature is approximately four times lower than the corresponding crack growth rate at reactor hot leg temperature (considering the standard thermal activation energy for crack growth of 130 kJ/mole per MRP-115). The PWHT applied to the cold leg locations is expected to result in similar residual stress levels as those calculated for the hot leg drain nozzle. Structural IntegrIty Associates, lnc Mr. William Sims RAM-14-008 February 25,2014 Page 5 of 16 through-wall over a relatively local circumferential region, ultimately resulting in detection of leakage prior to the possibility of unstable pipe rupture. These conclusions extend to the pipe connection Alloy 821182 full penetration weldments on the reactor cold legs at Palisades. The assessments presented above for the single hot leg location clearly bound the concern for PWSCC at each of the cold leg locations. The susceptibility to PWSCC initiation is greatly reduced for nickel-based weldments operating at reactor cold leg temperature, and the PWSCC growth rate at reactor cold leg temperature is approximately four times lower than the corresponding crack growth rate at reactor hot leg temperature (considering the standard thennal activation energy for crack growth of 130 kllmole per MRP-115). The PWHT applied to the cold leg locations is expected to result in similar residual stress levels as those calculated for the hot leg drain nozzle. lJ Structural Integrity Associates, Inc'"

Mr. William Sims RAM-14-008 February 25, 2014 Page 6 of 16 Figure 1. Finite Element Model Structural Integrity Associates, Inc Mr. William Sims RAM-14-008 1 Figure 1. Finite Element Model February 25, 2014 Page 6 of 16 e Slruclurallnlegrily Associates, Inc.-

Mr. William Sims RAM-14-008 February 25, 2014 Page 7 of 16 Structural Integrity Associates, Inc Figure 2. Weld Nuggets Mr. William Sims RAM-14-008 Figure 2. Weld Nuggets February 25, 2014 Page 7 of 16 e Structural Integrity Associates, Inc.e

Mr. William Sims RAM-14-008 February 25, 2014 Page 8 of 16 Structural Integrity Associates, Inc Figure 3. Patch Nuggets Mr. William Sims RAM-14-008 Figure 3. Patch Nuggets February 25, 2014 Page 8 of16 e Struclurallnlegrity Associates, Inc.-

Mr. William Sims RAM-14-008 February 25, 2014 Page 9 of 16

Figure 4. Radial Stresses at Operating Conditions StructuraI Integrity Associates, mc? Mr. William Sims RAM-14-008 Figure 4. Radial Stresses at Operating Conditions February 25, 2014 Page 9 of 16 e Structurallntegrily Associates, Inc.*

Mr. William Sims RAM-14-008 February 25, 2014 Page 10 of 16 Structural Integrity Associates, lnc 12.6344 .311488 1.Li74 26.2033 39.1492 6.16146 6.78443 19.7303 32.6762 45.6221 erating cycles Figure 5. Circumferential Stresses at Operating Conditions Mr. William Sims RAM-14-008 February 25, 2014 Page 10 of 16

6. 78443 1

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Structural Integrity Associates, Inc.-

Mr. William Sims RAM-14-008 February 25, 2014 Page 11 of 16 Crack Depth (in) Figure 6. Applied Stress Intensity vs. Crack Depth Circumferential Flaw Structural Integrity Associates, Inc. Deg 00 Deg3O Deg6O Deg9O 60 C.40 20 0 1 I I I I I 1/I t I

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Mr. William Sims RAM-14-008 February 25, 2014 Page 12 of 16 Deg 00 Deg 30 Deg 60 Deg 90 Time (yrs) Figure 7. Crack Growth for Circumferential Flaw Note: Wall Thickness is 4. Structural Integrity Associates, lnc C a. 0 I 0 100 200 300 Mr. William Sims RAM-14-008 4 I _____________ 1 ___________ _ _ ____________ L ____ I _______ _ I ____________ J ___________ _ 3 t-- ---!----I, c I -f -----

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Mr. William Sims RAM-14-008 February 25, 2014 Page 13 of 16 Crack Depth (in) L = 0.5 L=i L=2 Figure 8. Applied Stress Intensity vs. Crack Depth Axial Flaw StructuraI Integrity Associates, Inc 30 20 10 0 0 1 2 3 4 Mr. William Sims RAM-14-008 £: ! ~ fI!.. oM: - ~ 30------~----~------~----~----~------r_----~----_,

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Structural Integrity Associates, Inc.1e

Mr. William Sims RAM-14-008 February 25, 2014 Page 14 of 16 lime (yrs) L=O.5 L1 L2 100 Figure 9. Crack Growth for Axial Flaw Note: Wall Thickness is 4. Structural Integrity Associates, Inc 4 3 I 0 0 20 40 60 80 Mr. William Sims RAM-14-008 4r---~-----r----~----~----~--~----~----~----~--~ 3 I-I jf . / .c _* __ **.. _**_ :1 / ~ 13 . -t-~ -c 1 [ , rzmu[uuum,uuurumm:uuuuTumTmuu1 o~'----~----~----~----~----~----~----~----~----~----~ February 25,2014 Page 14 of 16 L= 0.5"

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--- L=2" o 20 40 60 80 100 llme (yrs) Figure 9. Crack Growth for Axial Flaw Note: Wall Thickness is 4". t:J Stluctulallnteglity Associates, Inc.e

Mr. William Sims RAM-14-008 February 25, 2014 Page 15 of 16 Figure 10. Extent of Partial-Arc Through-wall Circumferential Flaw Assumed for Limit Analysis (ASME Code, Level A Service Limits) Note: Flaw is Located at the Nozzle-Weld Interface. fStructuraI Integrity Associates, Inc ANSYS R14.5 FEB 24 2014 11:14:39 PLF NO. 1 Limit Analvsi.s Mr. William Sims RAM-I 4-008 1 EI.EM!NI'S Limit Analvsis February 25,2014 Page 15 ofl6 ANSYS ~14.S FEB 24 2014 11 :14 :39 PLOr ro. 1 Figure 10. Extent of Partial-Arc Through-wall Circumferential Flaw Assumed for Limit Analysis (ASME Code, Level A Service Limits) Note: Flaw is Located at the Nozzle-Weld Interface. lJ Structural Integrity Associates, Inc.-

Mr. William Sims RAM-14-008 February 25, 2014 Page 16 of 16 7.56858 2.42115 10.1423 4.99486 ANSYS R14.5 FEB 24 2014 13 : 17 :33 PIDINO. 1 Structural Integrity Associates, Inc ELEMENTS PPES 17.8634 12.716 15.2897 Limit Analysis 2.72628 .152568 5.3 Figure 11. Extent of Through-Wall Axial Flaw Assumed for Limit Analysis (ASME Code, Level A Service Limits) Mr. William Sims RAM-14-008 1 ELEMENTS PRES February 25, 2014 Page 16 ofl6 FEB 24 2014 13:17 :33 PIDI' ID. 1 -17.8634 -12.716 -7.56858 -2.42115 2.72628 -15.2897 -10.1423 -4.99486 .152568 5.3 Limit is Figure 11. Extent of Through-Wall Axial Flaw Assumed for Limit Analysis (ASME Code, Level A Service Limits) l) Structural Integrity Associates, Inc."}}

v t e Letter Sequence Request
TAC:MF3508, (Open)
Initiation Request, Request, Request, Request, Request Acceptance... Supplement, Supplement Administration Questions 27 February 2014 Answers 1 March 2014 Results Approval... Other: ML14223B226
MONTHYEAR2014-03-11 [Table View]

PNP 2014-015, Relief Request Number RR 4-18 - Proposed Alternative Use of Alternate ASME Code Case N-770-1 Baseline Examination

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PNP 2014-015, Relief Request Number RR 4-18 - Proposed Alternative Use of Alternate ASME Code Case N-770-1 Baseline Examination

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