ML070110359

Request to Extend the Second 10-Year, American Society of Mechanical EngineersSection XI, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations - Relief Request No. 34, Revision 1

ML070110359
Person / Time
Site:	Palo Verde
Issue date:	01/04/2007
From:	Mauldin D Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-05621-CDM/SAB/RJR
Download: ML070110359 (14)

Text

LAPS David Mauldin Vice President 10 CFR 50.55a(a)(3)(i)

Mail Station 7605 Palo Verde Nuclear Nuclear Engineering Tel: 623-393-5553 PO Box 52034 Generating Station and Support Fax: 623-393-6077 Phoenix, Arizona 85072-2034 102-05621-CDM/SAB/RJR January 04, 2007 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

References:

1. Letter from Nuclear Regulatory Commission to Westinghouse Electric Company, "Summary of teleconference with the Westinghouse Owners Group regarding potential one cycle relief of reactor pressure vessel shell weld inspections at pressurized water reactors related to WCAP-16168-NP, 'Risk-Informed Extension of Reactor Vessel In-Service Inspection Intervals'," dated January 27, 2005.

2. APS letter 102-05446-CDM/SAB/RJR, "Request to Extend the Second 10-Year, American Society of Mechanical EngineersSection XI, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations - Relief Request No. 34," dated May 4, 2006.

SUBJECT:

Palo Verde Nuclear Generating Station (PVNGS)

Units 2 and 3 Docket Nos. STN 50-5291530 Request to Extend the Second 10-Year, American Society of Mechanical EngineersSection XI, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations - Relief Request No. 34, Revision I This letter revises the previously submitted request, Reference 2, to correct Section 1.0 of the original enclosure to add Category B-A, Item B1.12 that was mistakenly omitted from Section 1.0. Justification for the proposed relief for examination category B-A was addressed in the original submittal. The specific weld identification numbers for Item No. B13.12 identified below were listed in the attachment to the original submittal enclosure.

Unit Weld Identification Numbers Unit 2 2-001-003, 004, 005, 007, 008, 009, 011, 012 and 013 Unit 3 3-001-003, 004, 005, 007, 008, 009, 011, 012 and 013 Ao04 7

NRC Document Control Desk Page 2 Relief Request No. 34 R1 The technical basis to extend the Second 10-year ISI interval to the end of the spring 2008 and spring 2009 refueling outages for Palo Verde Units 2 and 3, respectively, presented in the original submittal remain applicable. The five areas that form the basis for the technical justification are provided below and are addressed in the attached relief request submittal.

A. Plant-specific reactor pressure vessel (RPV) inservice inspection (ISI) history B. Pressurized-water reactor (PWR) RPV ISI history C. Degradation mechanisms in the RPV D. Material condition of the RPV relative to embrittlement E. Operational experience relative to RPV structural integrity challenging events Pursuant to 10 CFR 50.55a(a)(3)(i), Arizona Public Service Company (APS) is requesting approval to use an alternative to the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, Paragraph IWB-2412, Inspection Program B, for PVNGS Units 2 and 3. The proposed alternative requests relief to defer the reactor vessel weld examinations of Units 2 and 3 for one fuel cycle as described in the enclosure to this letter.

As stated in the Enclosure, this request does not apply to any dissimilar metal welds, including Alloy 600 base metal or Alloy 82/182 weld material where primary water stress corrosion cracking is a concern or any other augmented inspection requirements imposed. The technical justification to extend PVNGS's second inspection interval performance of Category B-A and B-D examinations by one fuel cycle is consistent with the guidance provided in Reference 1. APS's proposed extension of the inservice inspection interval for these examinations will continue to provide an acceptable level of quality and safety, as described in the enclosed relief request.

APS requests staff approva! by May 17, 2007. This letter contains no new commitments or revisions to existing commitments.

If you have any questions about this change, please telephone Thomas N. Weber at (623) 393-5764.

Sincerely, CDM/SAB/RJ RI

Enclosure:

Relief Request No. 34, Revision 1 - Request to Extend the Second 10-Year, American Society of Mechanical EngineersSection XI, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations cc: B. S. Mallett NRC Region IV Regional Administrator M. B. Fields NRC NRR Project Manager G. G. Warnick NRC Senior Resident Inspector for PVNGS

ENCLOSURE Relief Request No. 34, Revision 1 - Request to Extend the Second 10-Year, American Society of Mechanical. EngineersSection XI, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations

Enclosure - RR 34, Revision 1 Relief Request No. 34, Revision 1 Request to Extend the Second 10-Year Inservice Inspection Program Interval for Palo Verde Units 2, and 3 Reactor Vessel Weld Examinations 1.0 ASME Code Component(s) Affected The affected components are the PVNGS Unit 2 and 3 Reactor Vessels (RV),

specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI examination categories and item numbers. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.

Examination Item No Description Category B-A B1.11 Circumferential Shell Welds B-A BI.12 Longitudinal Shell Welds B-A B1.22 Meridional Shell Welds (Bottom Head only)

B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas (Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code Section Xl is referred to as "the Code")

2.0 Applicable Code Edition and Addenda

The PVNGS Unit 2 and 3 second 10-year Interval Inservice Inspection (ISI)

Program Plan is prepared to comply with the 1992 Edition 1992 Addenda of the Code.

3.0 Applicable Code Requirement

ASME Section XI - IWA-2432 - Inspection Program B states in part:

"The inspection intervals shall comply with the following, except as modified by IWA-2430(d):" "Successive Inspection Intervals - 10 years following the previous inspection interval."

ASME Section XI - IWA-2430(d) states:

"For components inspected under Program B, each of the inspection intervals may be extended or decreased by as much as 1 year. Adjustments shall not cause successive intervals to be altered by more than 1 year from the original pattern of intervals."

Page 1

Enclosure - RR 34, Revision 1 ASME Section Xl - IWB-2412(b) states:

"The inspection interval specified in IWB-2412(a) may be decreased or extended by as much as 1 year to enable an inspection to coincide with a plant outage, within the limitations of IWA-2430(d)"

ASME Section X1 - Table IWB-2500-1, Examination Category, B-A, requires a volumetric examination on all welds.

ASME Section X1 - Table IWB-2500-1, Examination Category, B-D, requires a volumetric examination on all nozzles.

4.0 Reason for Request

The intent of the request is to extend the ISI interval for Examination Category B-A and B-D by one refueling cycle beyond the currently scheduled inspection, to allow time for NRC review of industry efforts to extend the ISI interval for the subject examinations from 10 to 20 years. These efforts use ASME Section Xl Code Case N-691 (Reference 1) as a basis for using risk-informed insights to show that extending the inspection interval from 10 to 20 years results in a small change in the RV failure frequency that satisfies the requirements of Regulatory Guide 1.174 (Reference 2). Following NRC approval of these efforts, APS intends to submit a separate request to extend the current 10-year interval for PVNGS to 20 years.

The 20-year inspection interval will result in a reduction in man-rem exposure and examination costs.

5.0 Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(a)(3)(i), Arizona Public Service Company (APS) hereby requests approval to use an alternative to the requirements of the ASME Boiler and Pressure Vessel Code, Section Xl, Paragraph IWB-2412, Inspection Program B, for PVNGS Units 2 and 3. The alternative is to defer the Unit 2 and 3 reactor vessel weld examinations for one fuel cycle. The one cycle deferral will allow additional time for completing evaluations and staff review associated with Westinghouse Owners Group Topical Report, WCAP-16168 (Reference 3). The NRC has communicated to the Westinghouse Owners Group that the staff would agree to licensees submitting a one cycle relief request for an extension.

Currently, PVNGS Units 2 and 3 are in the third periods of their second ten-year ISI interval. The second ISI interval is currently scheduled to end on March 17, 2007, for Unit 2 and January 10, 2008, for Unit 3. Applying the extension allowed by IWA-2430(d)] would extend the end of the interval until March 17, 2008, for Unit 2 and January 10, 2009, for Unit 3. The 2R14 and 3R14 refueling outages are currently scheduled for the spring of 2008 and 2009, respectively. However, an additional amount of extension time will be required to capture the 2R1 4 and 3R1 4 outages, and the RV examinations, in their respective Second Intervals.

Page 2

Enclosure - RR 34, Revision 1 The additional extension being requested is less than 60 days for Unit 2 and less than 150 days for Unit 3. Although the proposed inspection dates are one refueling outage beyond the Code required inspection interval, the net duration between the reactor vessel inspections will not be more that the maximum eleven years allowed by the Code.

The guidance for the technical basis to extend the 10-year RV ISI interval by one refueling cycle are contained in a letter from R. Gramm of the NRC to G. Bischoff of the Westinghouse Owners Group, dated January 27, 2005 (Reference 4). The following information provides APS' technical justification that the current ISI interval can be extended while providing an acceptable level of quality and safety in accordance with 10 CFR 50.55a (a) (3) (i).

APS' technical basis for the relief request addresses the following five topics which were identified in the January 27, 2005 letter.

• Plant specific reactor vessel inservice inspection history.

" Fleetwide reactor vessel inservice inspection history.

" Degradation mechanisms in the reactor vessel.

• Material condition of the reactor vessel relative to embrittlement.

" Operational experience relative to RV structural integrity challenging events.

5.1 Reactor Vessel Inservice Inspection History for Units 2 and 3 PVNGS Unit 2 and 3 are in their second ISI interval for the reactor pressure vessel examinations. The preservice inspections (PSI) and one ISI have been performed on the Examination Category B-A and B-D welds to date.

The PSI was performed in accordance with ASME Section XI, 1974 Edition and Summer 1975 Addenda; and the ISI was performed in accordance with ASME Section XI, 1980 Edition, Winter 1981 Addenda, and Regulatory Guide 1.150 (Reference 5). The examinations performed in Units 2 and 3 have achieved acceptable coverage (i.e., >90% or examination of the maximum practical coverage). No reportable indications were found during these examinations. Based on the examination method and coverage obtained, it is reasonable to conclude that the examinations were of sufficient quality to detect any significant flaws that would challenge RV integrity. A detailed inspection history for the welds to which the subject examinations apply are contained tables 1 and 2 attached to this enclosure.

Page 3

Enclosure - RR 34, Revision 1 5.2 Fleetwide Reactor Vessel Inservice Inspection History As part of the technical basis for ASME Code Case N-691, a survey of RV ISI history for 14 pressurized water reactors was performed. These 14 plants represented 301 total years of service and included RVs fabricated by various vendors. The plants reported that no reportable findings had been discovered during examinations of Category B-A and B-D welds of their RVs.

It is widely recognized in the fracture mechanics community that fatigue crack growth of embedded flaws is substantially smaller than that of surface breaking flaws. PVNGS Units 2 and 3 contain one layer cladding. The completed cladding was 100% liquid penetrant (PT) examined during construction to assure freedom from lack of fusion or other linear defects open to the surface. This PT examination lowers the probability of surface breaking flaws propagating due to fatigue.

All Pressurized Water Reactor plants have performed their first 10-year ISI of the subject examinations. No surface-breaking or unacceptable near-surface flaws (i.e., defects) have been reported in any of these inspections performed per the requirements of Regulatory Guide 1.150, or ASME Section XI, Appendix VIII.

5.3 Degradation Mechanisms in the Reactor Vessel The welds for which the subject examinations are conducted are similar metal low alloy steel welds. The only currently known degradation mechanism for this type of weld is fatigue due to thermal and mechanical cycling from operational transients. Studies have shown that while flaw growth of simulated flaws in a reactor vessel would be small, the operational transient which has the greatest contribution to flaw growth is the cooldown transient.

Based on operating experience, the cooldown transient is a low frequency transient and is not expected to occur more than a few instances during the requested inspection extension period. Therefore, any flaw growth during the requested deferral period will be inherently small.

The fatigue usage factors for the welds in the subject examinations are much less than the ASME Code design limit of 1.0 after 40 years of operation.

These usage factors are calculated using a very conservative design duty cycle. It is very unlikely that more than a few of these events (e.g. heatup or cooldown) would actually occur during the extension period of this proposed alternative.

It is important to note that this request does not apply to any dissimilar metal welds, including Alloy 600 base metal or Alloy 82/182 weld material, where primary water stress corrosion cracking is a concern or any other augmented inspection requirements are imposed.

Page 4

Enclosure - RR 34, Revision 1 5.4 Material Condition of the Reactor Vessel Relative to Embrittlement The RV beltline is the limiting area in terms of embrittlement for the subject examinations. The composition of each material in the RV beltline, along with fluence and embrittlement data, can be found in the NRC Reactor Vessel Integrity Database (RVID) (Reference 6). This information is provided for PVNGS Units 2 and 3 in the following tables. Note: The RTPTS values have been updated as discussed below.

Palo Verde Unit 2 Material Values Contained in the RVID Major Material Region Description Cu Ni Un-Irradiated RTPTs Type Heat Location [wt%] [wt%] [OF] Method Plate F-765-6 Intermediate 0.040 0.670 10 Plant 78 Shell Specific Lower Plant Plate F-773-2 0.040 0.640 0 68 Shell Specific Weld MIL B-4 Axial Weld 0.090 0.040 -80 Plant 34 101-142A, B, C Specific Palo Verde Unit 3 Material Values Contained in the RVID Major Material Region Description Cu Ni Un-Irradiated RTPTS Type Heat Location ((O%] [wt%] [OF] Method Plate F-6407-5Sel Intermediate 0.050 0.610 -20 Plant Spcfc55 5 Shell Specific Plate LoerPlant 6 Plate Lower 0.040 0.650 0 Pecfi 68 2 Shell Specific Weld MIL B-4 Axial Weld 0.040 0.070 -50 Plant 30 101-142A, B, C I Specific I 10 CFR 50.61 currently provides pressurized thermal shock (PTS) screening criteria of RTPTs equal to 2701F for plates and axial welds and RTPTs equal to 300OF for circumferential welds. Based on current projections, the intermediate shell plate in Unit 2 is the most limiting material for PVNGS Unit 2 and 3. The projected RTPTS value of 78 0 F at 32 EFPY for this material is well below the PTS screening criteria. Furthermore, it is recognized by the NRC and industry that a large amount of conservatism exists in the current PTS screening criteria (Reference 7). In the NRC PTS Risk Re-evaluation, results have shown that it may be possible to remove an amount of conservatism equivalent to reducing a plant's RTPTS value by at least 70 0 F.

While the exact amount of conservatism that will be removed has not been determined, it is clear that PVNGS Unit 2 and 3 will be well below the current Page 5

Enclosure - RR 34, Revision 1 PTS screening criteria during the extension period and further below the potential revised PTS screening criteria.

5.5 Operational Experience Relative to RV Structural Integrity Challenging Events It is widely recognized that the greatest possible challenge to reactor pressure vessel integrity for a PWR is PTS. A PTS event can be generally described as a rapid cooling of the RV followed by a late repressurization. Plants (including PVNGS) have taken steps such as implementing emergency operating procedures (EOPs) and operator training to lower the likelihood of a PTS event occurring. Due to the implementation of such measures, industry experience indicates the number of occurrences of PTS events fleetwide is very small. When considered over the combined fleetwide PWR operating history, the frequency of PTS events is very small. When considering the frequency of PTS events and the length of the requested extension, the probability of a PTS event occurring during the requested extension is also very low. Combining the low probability of a PTS event with the low probability of a flaw existing in the RV (given the previously discussed inspection history), the probability of RV failure due to PTS is also very small.

PVNGS Units 2 and 3 have implemented EOPs and operator training to prevent the occurrence of PTS events. Consistent with the Combustion Engineering (CE) Emergency Response Guidelines (ERGs), the PVNGS EOPs allow operators to identify the onset of PTS conditions and provide the steps required to mitigate any cold pressurization challenge to RV integrity.

The basic PTS mitigation strategy of the PVNGS EOPs involves 1) termination of the primary system cooldown, 2) termination of emergency core cooling system flow (if proper criteria are met), 3) depressurization of the primary system, 4) establishment of stable primary system conditions in the normal operating range, and 5) implementation of a thermal "soaking" period prior to any cooldown outside of the normal operating region. By combining

1) the basic requirements of the CE ERGs, 2) the use of plant specific setpoints with a defined technical basis, and 3) the formal reconciliation of any differences between the CE ERG reference plant and PVNGS, the PVNGS EOPs provide adequate means for preventing potential PTS transients.

The current requirements for inspection of RV pressure-retaining welds have been in effect since the 1989 Edition of ASME Code, Section Xl. The industry has expended significant cost and man-rem exposure that have shown no service-induced flaws in the ASME Section XI Examination Category B-A or B-D RV welds. ASME Section XI, Code Case N-691 and industry efforts have shown that risk insights can be used to extend the reactor vessel ISI interval from 10 to 20 years. The 10-year extension satisfies the change in risk requirements of Regulatory Guide 1.174; and, in accordance with 10 CFR 50.55a (3) (i), maintains an acceptable level of quality and safety. Based on these efforts having shown that the risk of vessel failure with a 10-year Page 6

Enclosure - RR 34, Revision 1 inspection interval extension is low and achieves an acceptable level of quality and safety, it is reasonable to conclude that one refueling cycle extension will also achieve an acceptable level of quality and safety. On the basis of the above discussion, the risk associated with extending the inspection interval by one refueling cycle is small. Therefore, APS considers the proposed alternative for the subject examinations at PVNGS Units 2 and 3 to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(3)(i).

6.0 Duration of Proposed Alternative The proposed alternative requested would extend the second ISI interval of PVNGS Unit 2 and 3 for one refueling cycle and would apply to the Examination Category B-A and B-D RV welds beyond the ASME Code required 10-year inspection interval. This request is applicable to the second inspection interval only. If this relief request is approved, the second ISI interval will end at the conclusion of the spring 2008 and 2009 (2R1 4 and 3R1 4) outages for the subject examinations.

7.0 References

1. ASME Boiler and Pressure Vessel Code, Code Case N-691, "Application of Risk-Informed Insights to Increase the Inspection Interval for Pressurized Water Reactor Vessels,"Section XI, Division 1, November 2003.

2. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.

3. Westinghouse Owners Group Topical Report, WCAP-16168-NP, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval,"

December 2005.

4. R. Gramm of the NRC to G. Bischoff of the WOG, "Summary of Teleconference with the Westinghouse Owners Group Regarding Potential One Cycle Relief of Reactor Pressure Vessel Shell Weld Inspections at Pressurized Water Reactors Related to WCAP-16168- NP, 'Risk-Informed Extension of Reactor Vessel In-Service Inspection Intervals,"' dated January 27, 2005.

5. Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds during Preservice and Inservice Examinations," February 1983.

6. Nuclear Regulatory Commission Reactor Vessel Integrity Database, dated July 22, 1995.

Page 7

Enclosure - RR 34, Revision 1

7. NRC Memorandum, Thadani to Collins, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Criteria in the PTS Rule (10 CFR 50.61)," dated December 31, 2002.

8.0 Precedent Palisades Nuclear Plant November 29, 2005 ML053200296 Indian Point, March 16, 2006 ML060740187 Nuclear Generating Unit No. 2 Sequoyah Nuclear Plant, February 3, 2006 ML060100080 Unit 1 Page 8

Attachment Attachment Table 1 Palo Verde Unit 2 Inservice Inspection Results Weld ID ASME Date Last  % Coverage obtained # of # of indications Growth of Weld Inspected reportable currently being indications Category indications* monitored* currently being monitored* (in) 2-001-001 B-A 09/23/97 22% ** None None None 2-001-002 B-A 09/23/97 90.2% None None None 2-001-003 B-A 09/23/97 100% None None None 2-001-004 B-A 09/24/97 100% None None None 2-001-005 B-A 09/23/97 100% None None None 2-001-006 B-A 09/24/97 93% None None None 2-001-007 B-A 09/22/97 95% None None None 2-001-008 B-A 09/22/97 95% None None None 2-001-009 B-A 09/22/97 95% None None None 2-001-010 B-A 09/21/97 91.2% None None None 2-001-011 B-A 09/20/97 100% None None None 2-001-012 B-A 09/20/97 100% None None None 2-001-013 B-A 09/19/97 100% None None None 2-001-014 B-A 09/20/97 92.6% None None None 2-001-016 B-D 09/22/97 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld. ***

2-001-016-IR B-D 09/24/97 92% None None None 2-001-017 B-D 09/23/97 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld. ***

2-001-017-IR B-D 09/24/97 92% None None None 2-001-019 B-D 09/22/97 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld.

• 2-001-019-IR B-D 09/24/97 92% None None None 2-001-020 B-D 09/22/97 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld.

2-001-020-IR B-D 09/24/97 92% None None None Notes:

• Due to improvement in inspection technology, the most recent inspection is considered to be of the greatest quality of the inspections performed. Therefore, the inspection data provided in this table is for the most recent inservice inspection.

• • Coverage of the lower head meridional weld is limited by interference with the flow skirt and ICI nozzles.

Coverage of the nozzle to vessel circumferential weld scans limited by saddle geometry and nozzle boss interference.

Page 1

Attachment Table 2 Palo Verde Unit 3 Inservice Inspection Results Weld ID ASME Date Last  % Coverage obtained # of # of indications Growth of Weld Inspected reportable currently being indications Category indications* monitored* currently being monitored* (in) 3-001-001 B-A 10/01/98 20% ** None None None 3-001-002 B-A 09/30/98 90.6% None None None 3-001-003 B-A 10/02/98 100% None None None 3-001-004 B-A 10/02/98 100% None None None 3-001-005 B-A 10/02/98 100% None None None 3-001-006 B-A 10/02/98 93% None None None 3-001-007 B-A 10/02/98 95% None None None 3-001-008 B-A 10/02/98 95% None None None 3-001-009 B-A 10/02/98 95% None None None 3-001-010 B-A 10/01/98 91.2% None None None 3-001-011 B-A 09/30/98 100% None None None 3-001-012 B-A 10/01/98 100% None None None 3-001-013 B-A 10/01/98 100% None None None 3-001-014 B-A 09/30/98 92% None None None 3-001-016 B-D 10/03/98 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld. *'*

3-001-016-IR B-D 10/03/98 90% *** None None None 3-001-017 B-D 10/03/98 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld. ***

3-001-017-IR B-D 10/03/98 90% *** None None None 3-001-019 B-D 10/02/98 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld. ***

3-001-019-IR B-D 10/02/98 90% *** None None None 3-001-020 B-D 10/02/98 96.5% coverage for reflectors None None None parallel to weld. 80.1%

coverage for reflectors transverse to weld. ***

3-001-020-IR B-D 10/03/98 90% None None None Notes:

• Due to improvement in inspection technology, the most recent inspection is considered to be of the greatest quality of the inspections performed. Therefore, the inspection data provided in this table is for the most recent inservice inspection.

• • Coverage of the lower head meridional weld is limited by interference with the flow skirt and ICI nozzles.

Coverage of the nozzle to vessel circumferential weld and inner radius scans limited by nozzle saddle geometry and boss interference.

Page 2