ML050970267
| ML050970267 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/31/2005 |
| From: | Domonique Malone Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML050970267 (12) | |
Text
ASPA NM C 7 Palisades Nuclear Plant Committed to Nuclear Exce~lflence Paiaeula ln COperated by Nuclear Management Company, LLC March 31, 2005 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 Request for Authorization to Extend the Third 1 0-Year ISI Interval for Reactor Vessel Weld Examination
References:
- 1)
Westinghouse Owners Group Topical Report, WCAP-16168-NP, uRisk-Informed Extension of Reactor Vessel Inservice Inspection Interval," dated October 2003
- 2)
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-Senrice Inspection Intervals," dated January 27, 2005
- 3)
Letter from NMC to NRC, "Request for Authorization to Extend the Third 10-Year ISI Interval for Reactor Vessel Visual Examination," dated March 31, 2005
- 4)
Letter from NMC to NRC, 'Request for Authorization to Extend the Third 10-Year IST Interval for Certain Relief Valves," dated March 31, 2005 Pursuant to 10 CFR 50.55a(a)(3)(i), Nuclear Management Company, LLC (NMC) is requesting Nuclear Regulatory Commission (NRC) approval for the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, paragraph IWB-2412, Inspection Program B, for the Palisades Nuclear Plant. NMC is submitting this relief request because the Westinghouse Owners Group Topical Report (Reference 1) is not currently being reviewed and therefore the NRC 'agreed that licensees could submit a one-cycle relief request.
27780 Blue Star Memorial Highway
- Covert, Michigan 49043-9530 4w<7 Telephone: 269.764.2000 A
Document Control Desk Page 2 Palisades is currently in the third inspection interval. The third inspection interval began on May 12, 1995, and considering the ASME Code-allowed extensions, will end on December 12, 2006. The examination of the reactor vessel welds (Category B-A), the nozzle-to-vessel welds and inner radius sections (Category B-D), and reactor vessel nozzle-to-piping welds (Category B-J), for the third interval is currently scheduled during the spring 2006 refueling outage. As a result of the adoption of a Risk Informed Inservice Inspection Program at Palisades, the Category B-J welds are presently included in the augmented inspection program as defense-in-depth exams.
NRC approval is requested to extend the third inspection interval for the Category B-A, B-D, and B-J welds for one refueling cycle. The technical justification for this request is consistent with the guidance provided in Reference 2. The extension of the inspection interval for these examinations will still result in an acceptable level of quality and safety, as described in the enclosed request.
This request is associated with two other requests, both by letter dated March 31, 2005 (Reference 3, Reference 4). Both Reference 3 and Reference 4 were to align with the full core off load during the fall 2007 refueling outage. Approval of Reference 3 and Reference 4 is contingent upon the approval to extend the third 10-Year ISI interval for the reactor vessel weld examination.
NMC requests approval by March 1, 2006, however, NMC would like approval sooner to accommodate outage planning.
Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments.
Daniel J. Malone Site Vice President, Palisades Nuclear Plant Nuclear Management Company, LLC Enclosures (1)
Attachment (1)
CC Administrator, Region l1l, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC
ENCLOSURE 1 REQUEST FOR AUTHORIZATION TO EXTEND THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL FOR REACTOR VESSEL WELD EXAMINATION PALISADES NUCLEAR PLANT 1.0 ASME Code Component(s) Affected The affected component is the Palisades Nuclear Plant reactor vessel (RV), specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section Xl examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code, Section Xl.
Examination CateQorv Item No.
Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas B-J*
B9.11 Circumferential Welds in Piping
- As a result of the adoption of a Risk Informed Inservice Inspection Program at Palisades, the Category B-J welds are presently included in the augmented inspection program as defense-in-depth exams.
(Throughout this request, the above examination categories are referred to as uthe subject examinations," and the ASME BPV Code,Section XI, is referred to as "the Code.")
2.0
Applicable Code Edition and Addenda
The Palisades Nuclear Plant third interval Inservice Inspection (ISI) Program Plan is prepared to the 1989 Edition of the Code.
3.0
Applicable Code Requirement
IWB-2412, Inspection Program B, requires volumetric examination of essentially 100%
of RV pressure retaining welds identified in Table IWB-2500-1, once each ten-year interval. In accordance with iWA-2430(d) and IWA-2430(e), Palisades third inspection interval is currently scheduled to conclude on or before December 12, 2006.
Page 1 of 7
4.0
Reason for Request
An alternative is requested from the requirement of IWA-2412, Inspection Program B, that volumetric examination of essentially 100% of RV pressure retaining welds, examination categories B-A, B-D and B-J, be performed once each ten-year interval.
Extension of the inspection interval, for examination category B-A, B-D and B-J, by one refueling cycle beyond the currently scheduled inspection is requested.
The intent of the requested one refueling cycle extension is to allow for deferment of the subject examinations 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 4), as a basis for using risk-informed insights to show that extending the inspection interval from 10 to 20 years results in a change in RV failure frequency that satisfies the requirements of Regulatory Guide 1.174 (Reference 7). Following NRC approval of these efforts, NMC intends to submit a separate request to extend the current 10-year interval for Palisades Nuclear Plant to 20 years.
5.0 Proposed Alternative and Basis for Use The third inspection interval for Palisades started on May 12,1995, and will end on or before December 12, 2006. This inspection interval includes credit for the IWA-2430(d) allowed one-year extension and the IWA-2430(e) allowed 215-day extension, due to the 2001 extended maintenance outage. The subject examinations are currently scheduled during the spring 2006 refueling outage. The proposed inspection date is one refueling cycle beyond the Code-allowed inspection interval. In accordance with 10 CFR 50.55a(a)(3)(i), this interval extension is requested on the basis that the current inspection interval can be extended, while providing an acceptable level of quality and safety.
The requirements for a technical basis to extend the 10-year RV ISI interval by one refueling cycle are contained in a letter to the Westinghouse Owners Group, dated January 27, 2005 (Reference 3). This letter provides the basis for the one refueling cycle extension of the 10-year inspection interval for the subject examinations.
The technical justification for the extension of the inspection interval for the subject examinations was developed based on the guidance provided in Reference 3. The technical justification consists of five areas. These are:
5.1 -
Plant specific RV ISI history 5.2 Fleetwide RV ISI history 5.3 Degradation mechanisms in the RV 5.4 Material condition of the RV relative to embrittlement 5.5 Operational experience relative to RV structural integrity challenging events Page 2 of 7
5.1 Palisades Reactor Vessel Inservice Inspection History Palisades is in its third ISI interval for the RV. Two inservice inspections have been performed on the Category B-A, B-D and B-J welds to date. In summary, these inspections have been performed in accordance with Regulatory Guide 1.150 (Reference 8), and have achieved acceptable coverage, with no reportable indications found. 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 of the subject examinations is contained in Attachment 1.
The welds connecting the primary coolant system hot and cold leg loop piping to the RV nozzles are classified as Category B-J welds, in accordance with the ASME Code, Section Xi, 1989 Edition. However, as a result of the adoption of a Risk Informed Inservice Inspection Program at Palisades, these welds are presently included in the augmented inspection program as defense-in-depth exams. These welds were last inspected in 1995, and the results of these exams are included in. By letter dated March 1, 2002 (Reference 9), Palisades submitted to the NRC the Risk Informed Inservice Inspection Program. In that submittal, Palisades committed to continue to inspect these welds as part of the ASME Code, Section Xl, RV inspection program. By letter dated May 19, 2003 (Reference 11),
the NRC issued the safety evaluation approving the Risk Informed Inspection Program. The inspection of these welds is tied to the inspection interval associated with the B-A and B-D welds. Therefore, these welds are included in this relief request The segments connecting the primary coolant system hot and cold leg loop piping to the RV nozzles were ranked as low safety significant by the expert panel as part of the risk ranking process. Additionally, these segments contributed less than 0.01%
of the system total piping segment core damage frequency. Therefore, the impact on the delta risk evaluation was inconsequential. Changing the inspection interval for these welds would have no effect on the conclusions in the analyses. These welds will be inspected during the next mechanized RV examination.
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 (PWRs) was performed. These 14 plants represented 301 total years of service, and included RVs fabricated by various vendors. These plants reported that no reportable findings had been discovered during examinations of their RVs category B-A, B-D, and B-J welds.
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. Surface breaking flaws in the RV cladding are typically a result of lack of fusion defects between bands of cladding. In studies performed by Pacific Northwest National Laboratory for the NRC Pressurized Thermal Shock (PTS) Risk Page 3 of 7
Reevaluation, it was determined that in plants with multi-pass cladding, for a flaw to exist through the cladding, two flaws would have to be aligned on top of one another; The probability of this occurring is very low (<.0001). The Palisades RV is constructed with multi-pass cladding, and therefore, has a low probability of containing through-cladding surface-breaking flaws.
All PWR plants, except one, have performed their first 10-year ISI of the subject examinations. No surface-breaking or near-surface flaws of any significance have been found in any of these inspections performed per the requirements of Regulatory Guide 1.150 or ASME Section Xl, 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 RV would be small, the operational transient which has the greatest contribution to flaw growth is the cooldown transient. The cooldown transient is a low frequency transient, and is not expected to occur more than once 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 basemetal, or Alloy 82/182 weld material where primary water stress corrosion cracking is a concern.
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 RV Integrity Database (RVID). This information is provided for Palisades in the table below.
Page 4 of 7
Palisades-Specific Material Values Drawn from the RVID Major Material Region DescriptionU Ni P
Un-Irradiated RTNDT RTps Type ID Location
[wt%]
[wt%l
[wt%]
J0F]
Method
@EOL 1
Axial Weld 3-112A lower 0.213 1.010 0.019
- 56 Generic 268.6 2
Axial Weld 3-112B lower 0.213 1.010 0.019
- 56 Generic 268.6 3
Axial Weld 3-112C lower 0.213 1.010 0.019
- 56 Generic 268.6 4
Axial Weld 2-112A upper 0.213 1.010 0.019
- 56 Generic 268.6 5
Axial Weld 2-112B upper 0.213 1.010 0.019
- 56 Generic 268.6 6
Axial Weld 2-112C upper 0.213 1.010 0.019
- 56 Generic 268.6 7
Circ Weld 9-112 intermediate 0.203 1.018 0.013
- 56 Generic 281.5 8
Plate D3804-1 lower 0.190 0.480 0.016 0
ASME NB-2331 187.3 9
Plate D3804-2 lower 0.190 0.500 0.015
-30 MTEB 5-2 159.9 10 Plate D3804-3 lower 0.120 0.550 0.010
-25 MTEB 5-2 106.6 11 Plate D3803-1 upper 0.240 0.510 0.009
-5 ASME NB-2331 194.4 12 Plate D3803-2 upper 0.240 0.520 0.010
-30 MTEB 5-2 194.9 13 Plate D3803-3 upper 0.240 0.500 0.011
-5 ASME NB-2331 194.4 10 CFR 50.61 currently provides PTS screening criteria of RTPTs equal to 270TF for plates and axial welds, and RTPTs equal to 3000F for circumferential welds. For Palisades, the axial welds are the limiting material, and their RTpTs value at end of life (EOL) approaches the current PTS screening criteria. However, it is recognized by the NRC and industry that a large amount of conservatism exists in the current PTS screening criteria. 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 700F. While the exact amount of conservatism that will be removed has not been determined, it is clear that Palisades will be below the current PTS screening criteria during the extension period, and further below the potential revised PTS screening criteria.
5.5 Operational Experience Relative to Reactor Vessel Structural Integrity Challenging Events It is widely recognized that the greatest possible challenge to RV integrity for a PWR is PTS. A PTS event can be generally described as a rapid cooling of the RV followed by late repressurization. Plants 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, 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, the probability of RV failure due to PTS is very small.
Page 5 of 7
Palisades has implemented EOPs and operator training to prevent the occurrence of PTS events. Palisades EOPs include caution statements at critical locations warning the operator of the potential for causing PTS.
Palisades has not performed an analysis in accordance with the requirements of Regulatory Guide 1.154 (Reference 10). Palisades minimizes the amount of neutron fluence accumulated at the RV beltline using a low leakage core, to keep the RV below the PTS screening criterion, obviating the need to perform this analysis.
There is a significant reduction in risk if the safety injection water temperature is increased. In an effort to minimize plant risk, a Palisades system operating procedure was revised stating the "...preferred [safety injection refueling water tank]
SIRWT temperature band is 850F to 90°F," and "...the SIRWT should be maintained greater than or equal to 800F whenever the PCS is in Mode 1, 2, 3, or 4."
The current requirements for inspection of RV pressure-containing welds have been in effect since the 1989 Edition of the Code. The industry has expended significant cost and man-rem exposure that have shown no service-induced flaws in the RV for ASME Section Xl, Category B-A, B-D, or B-J, RV welds. ASME Section XI Code Case N-691 and industry efforts have shown that risk-insights can be used to extend the RV inservice inspection interval from 10 to 20 years. This 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 inspection interval extension is low and achieves an acceptable level of quality and safety, it is reasonable to conclude that a one refueling cycle extension will also achieve an acceptable level of quality and safety. Furthermore, Section 5 provides a qualitative basis that the risk associated with extending the inspection interval by one refueling cycle is small. Therefore, NMC considers the proposed alternative for the subject examinations at Palisades 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 alternative is requested to extend the third ISI interval by one refueling cycle beyond the ASME Code required 10-year inspection interval, the Code-allowed twelve month extension, and the Code-allowed 215-day extension for the subject examinations. This request is applicable to the third inspection interval only. If this relief request is approved, the third ISI interval will end at the conclusion of the fall 2007 refueling outage for the subject exams.
Page 6 of 7
7.0 References
- 1. WCAP-1 6168-NP, 'Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," October 2003.
- 2. NRC to WOG, 'WOG Request for the Staff Review of Topical Report WCAP-1 6168-NP URisk-informed Extension of Reactor Vessel In-Service Inspection Intervals," August 18, 2004.
- 3. NRC to 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,"
January 27, 2005.
- 4. 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.
- 5. NRC Memorandum, Thadani to Collins, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Criteria in the PTS Rule (IOCFR50.61)," December 31, 2002.
- 6. NRC Reactor Vessel Integrity Database, Version 2.0.1, July 6, 2000.
- 7. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"
dated November 2002.
- 8. Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," dated February 1983.
- 9. NMC to NRC, "Relief Request: Alternate ASME Code,Section XI, Risk-informed Inservice inspection Program," dated March 1, 2002.
- 10. Regulatory Guide 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," dated January 1987.
11.NRC to NMC,"Palisades Plant - Risk-Informed Inservice Inspection Program (TAC NO. MB4420)," dated May 19, 2003.
Page 7 of 7
Attachment I Palisades Inservice Inspection Results Number of Growth of ASME ASME Percent Number of Indications Indications Weld Code Date Last Coverage Reportable Currently Being Currently Being Weld ID Category Item Inspected Obtained Indications*
Monitored*
Monitired*(in)
RPV Circumferential Weld 8-112 B-A B1.11 June 1995 90.37 0
0 N/A RPV Circumferential Weld 9-112 B-A B1.11 June 1995 90.37 0
0 N/A RPV Circumferential Weld 10-112 B-A B1.11 June 1995 100 0
0 N/A RPV Longitudinal Weld 1-1 12A B-A B1.12 June 1995 93.8 0
0 N/A RPV Longitudinal Weld 1-112B B-A B1.12 June 1995 93 0
0 N/A RPV Longitudinal Weld 1-112C B-A B1.12 June 1995 93 0
0 N/A RPV Longitudinal Weld 2-112A B-A 81.12 June 1995 97.86 0
0 N/A RPV Longitudinal Weld 2-112B B-A B1.12 June 1995 100 0
0 N/A RPV Longitudinal Weld 2-112C B-A B1.12 June 1995 100 0
0 N/A RPV Longitudinal Weld 3-112A B-A B1.12 June 1995 100 0
0 N/A RPV Longitudinal Weld 3-112B B-A B1.12 June 1995 100 0
0 N/A RPV Longitudinal Weld 3-112C B-A B1.12 June 1995 100 0
0 N/A RPV Circumferential Weld 4-113 B-A B1.21 June 1995 59.45 0
0 N/A RPV Closure Head Weld 1-118A B-A B1.22 Sept 1983 100 0
0 N/A RPV Closure Head Weld 1-118B B-A 81.22 Sept1983 100 0
0 N/A RPV Closure Head Weld 1-118C B-A 81.22 Sept 1983 100 0
0 N/A RPV Closure Head Weld 1-118D B-A B1.22 Sept 1983 100 0
0 N/A RPV Closure Head Weld 1-118E B-A B1.22 Sept 1983 100 0
0 N/A RPV Closure Head Weld 1-118F B-A B1.22 June 1995 100 0
0 N/A RPV Meridional Weld 1-113A B-A B1.22 June 1995 47 0
0 N/A RPV Meridional Weld 1-113B B-A B1.22 June 1995 53 0
0 N/A I of 3
Attachment I Palisades Inservice Inspection Results Number of Growth of ASME ASME Percent Number of Indications Indications Weld Code Date Last Coverage Reportable Currently Being Currently Being Weld ID Category Item Inspected Obtained Indications*
Monitored*
Monitired*(in)
RPV Meridional Weld 1-113C B-A B1.22 June 1995 53 0
0 N/A RPV Meridional Weld 1-113D B-A B1.22 June 1995 47 0
0 N/A RPV Meridional Weld 1-113E B-A 81.22 June 1995 53 0
0 N/A RPV Meridional Weld 1-113F B-A B1.22 June 1995 53 0
0 N/A RPV Circumferential Weld 7-112 B-A B1.30 June 1995 100 0
0 N/A RPV Closure Head to Flange Weld 6-118A B-A B1.40 1999/2001 67 0
0 N/A RPV Nozzle Inside Radius Weld 5-114A-IRS B-D B3.100 June 1995 100 0
0 N/A RPV Nozzle Inside Radius Weld 5-1148-IRS B-D B3.100 June 1995 100 0
0 N/A RPV Nozzle Inside Radius Weld 5-114C-IRS B-D B3.100 June 1995 100 0
0 N/A RPV Nozzle Inside Radius Weld 5-114D-IRS B-D B3.100 June 1995 100 0
0 N/A RPV Nozzle Inside Radius Weld 5-114E-IRS B-D B3.100 June 1995 100 0
0 N/A RPV Nozzle Inside Radius Weld 5-114F-IRS B-D 3
B3.100 June 1995 100 0
0 N/A RPV Nozzle to Shell Weld 5-114A B-D B3.90 June 1995 100 0
0 N/A RPV Nozzle to Shell Weld 5-114B B-D B3.90 June 1995 100 0
0 N/A RPV Nozzle to Shell Weld 5-114C B-D B3.90 June 1995 100 0
0 N/A RPV Nozzle to Shell Weld 5-114D B-D B3.90 June 1995 100 0
0 N/A RPV Nozzle to Shell Weld 5-114E B-D B3.90 June 1995 100 0
0 N/A RPV Nozzle to Shell Weld 5-114F B-D B3.90 June 1995 100 0
0 N/A PCS-30-RCL-1A-15 Elbow to Transition Piece B-J B9.11 June 1995 100 0
0 N/A PCS-30-RCL-1A-16 Transition Piece to Nozzle B-J 89.11 June 1995 100 0
0 N/A PCS-30-RCL-1B-13 Elbow to Transition Piece B-J B9.11 June 1995 100 0
0 N/A PCS-30-RCL-1B-14 Transition Piece to Nozzle B-J B9.11 June 1995 100 0
0 N/A 2 of 3
Attachment I Palisades Inservice Inspection Results Number of Growth of ASME ASME Percent Number of Indications Indications Weld Code Date Last Coverage Reportable Currently Being Currently Being Weld ID Category Item Inspected Obtained Indications*
Monitored*
Monitired*(in)
PCS-30-RCL-2A-14 Elbow to Transition Piece B-J B9.11 June 1995 100 0
0 N/A PCS-30-RCL-2A-15 Transition Piece to Nozzle B-J B9.11 June 1995 100 0
0 N/A PCS-30-RCL-2B-14 Elbow to Transition Piece B-J B9.11 June 1995 100 0
0 N/A PCS-30-RCL-2B-15 Transition Piece to Nozzle B-J B9.11 June 1995 100 0
0 N/A PCS-42-RCL-1H-1 Nozzle to Transition Piece B-i B9.11 June 1995 100 0
0 N/A PCS-42-RCL-1H-2 Transition Piece to Pipe B-i B9.11 June 1995 100 0
0 N/A PCS-42-RCL-2H-1 Nozzle to Transition Piece B-3 B9.11 June 1995 100 0
0 N/A PCS-42-RCL-2H-2 Transition Piece to Pipe B-J B9.11 June 1995 100 0
0 N/A
- Note: Due to improvements in inspection technology, the most recent inspection is considered to be of the greatest quality of the inspections performed. In some instances, indications were found during inspections and then, in later inspections with improved equipment, were determined to be reflections rather than indications.
Therefore, the inspection data provided in this table is for the most recent inservice inspection.
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