CPSES-200700169, Inspection and Mitigation of Alloy 82/182 Pressurizer Butt Welds
| ML070320129 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 01/30/2007 |
| From: | Madden F TXU Generation Co, LP, TXU Power |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CPSES-200700169, Log TXX-07013 | |
| Download: ML070320129 (45) | |
Text
CTXU Power "XU Pcwer Mike Blevins Com, arche Peak Steam Seniur Vice President &
Elec-ýrc Station Chief Nuclear Officer P. 0- Box IG02 (EO1,)
Glen Pose,. TX 76043 Telt 254 897 5209 Fax-: 254 8a7 6652 mike-blevins@txu.com CPSES-200700169 Log # TXX-07013 January 30, 2007 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 INSPECTION AND MITIGATION OF ALLOY 82/182 PRESSURIZER BUTT WELDS REF: NEI Letter, "Industry Actions Associated with Potential Generic Implications of Wolf Creek Inspection Findings," from Marvin Fertel to Luis A. Reyes dated January 26, 2007
Dear Sir or Madam:
In October of 2006, while performing inspections of its Pressurizer Alloy 82/182 butt welds in accordance with the industry Materials Reliability Program MRP- 139, a PWR licensee discovered several circumferential indications in its Pressurizer surge, safety and relief nozzles. Because of the importance of this issue, I am submitting this letter to commit to the following actions planned for inspection or mitigation of Alloy 82/182 butt welds on pressurizer spray, surge and relief lines at Comanche Peak Steam Electric Station (CPSES) Units I and 2.
Inspection and mitigation activities of Pressurizer Alloy 82/182 butt welds at CPSES Unit I are scheduled to be completed during the refueling outage in Spring 2007.
Details concerning the CPSES Unit I inspection and mitigation activities are provided in Attachment 1. Future inspections of Pressurizer butt welds at CPSES Unit 1 will be performed in accordance with industry guidance (MRP-139). The results of future inspections or mitigations of Pressurizer Alloy 82/182 butt weld locations will be reported to the NRC within 60 days of startup from the outage during which they were performed.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance X110 Callaway
- Comanche Peak
- Diablo Canyon
- Palo Verde
- South Texas Project
- Wolf Creek
TXX-07013 Page 2 of 4 Inspection and mitigation activities of Pressurizer Alloy 82/182 butt welds at CPSES Unit 2 are scheduled to be completed during the refueling outage which is scheduled to begin March 29, 2008. Details concerning the CPSES Unit 2 inspection and mitigation activities are provided in Attachment 2. The schedule date is justified based on the infon-nation provided in the referenced letter and Attachment 3. Future inspections of pressurizer butt welds at Comanche Peak Unit 2 will be performed in accordance with industry guidance (MR.P-139). The results of future inspections or mitigations of Pressurizer Alloy 82/182 butt weld locations will be reported to the NRC within 60 days of startup from the outage during which they were performed.
The industry "Materials Reliability Program: Primary, System Piping Butt Weld Inspection and Evaluation Guidelines" ( MRP-139) was issued in September 2005.
Subsequent to that, vendors began developing programs for structural weld overlay.
Due to developing structural weld overlay processes, the availability of qualified contractors and equipment, and the long planning cycles necessary to prepare for refueling outages, CPSES personnel decided to schedule the inspection and mitigation of Pressurizer Alloy 82/182 butt welds at CPSES Unit 2 in the outage scheduled in Spring 2008. Compared to the U.S. PWR fleet, CPSES Unit 2 is the second lo*west unit in service years. We concluded that a three month delay of performing mitigation in Spring 2008 would have no safety impact and would allow the vendor's program to become more robust, reliable, and available. Comanche Peak Unit 2 is preparing a deviation justification for perfonning Unit 2 mitigation activities after December 3 1, 2007 in accordance with MRP- 139.
In addition to the inspection and mitigation actions described above, enhancements for monitoring primary system leakage are being used at Comanche Peak Units I and
- 2. These are described in Attachment 3. This enhanced guidance provides additional assurance that structural integrity is maintained and that any primary system pressure boundary leakage is discovered in a timely manner.
An effort is underway by the PWR Owners Group to standardize the guidance for leakage monitoring programs used by PWR plants. This program is under review and is expected to be finalized by mid-February 2007. CPSES will provide an update to the NRC as necessary regarding any changes to our leakage monitoring program by March 3 1, 2007.
Comanche Peak will also evaluate the feasibility of plant modifications to install diverse leakage detection capability. This may include, but is not limited to, video monitoring of Pressurizer piping, acoustic monitoring in the area of the Pressurizer, sensitive humidity monitoring, and other methods currently under evaluation. Plans for any additional capability which reliably and meaningfully adds to our ability to
TXX-07013 Page 3 of 4 diagnose primary system leakage, as well as installation schedules as appropriate will be submitted to the NRC by May 31, 2007.
If Comanche Peak Units I or 2 should shut down prior to presently planned inspection! mitigation outages due to unacceptable primary system leakage, and if the leakage cannot be confirmed to originate from a source other than the Pressurizer, a bare metal visual examination of Alloy 82/182 butt weld locations on the Pressurizer will be performed to determine whether the leakage originated at those locations.
Comanche Peak will continue to work with the industry to develop additional justification for inspection and mitigation past December 31, 2007. Additionally, we will accelerate the Unit 2 outage scheduled in March 2008 if analytical results do not demonstrate to the NRC that the current schedule is adequate. The Unit 2 schedule could also be accelerated if new information is obtained during upcoming inspections that challenge cutrent assumptions.
The NRC will be informed prior to any revision of the information contained in this letter.
This communication contains new licensing basis commitments for CPSES Units I and 2 which are delineated in Attachment 4.
Our staff is available to meet with the NRC to discuss any of the information in this letter. If there are any questions, please contact Fred Madden at (254)897-8601 or John Meyer at (254)897-6708.
Sincerely, TXU Generation Company LP By: TXU Generation Management Company LLC Its General Partner Mike Blevins By: *,/ "-?2I
,/Fr-e" W. Madden Director, Oversight and Regulatory Affairs
TXX-07013 Page 4 of 4 JCH Attachments: 1. Inspection and Mitigation Summary for Comanche Peak Unit 1
- 2. Inspection and Mitigation Summary for Comanche Peak Unit 2
- 3. Justification for Performing Unit 2 Mitigation After December 3 1, 2007
- 4. Licensing Basis Commitments for Units 1 and 2 c - James E. Dyer, Director, Office of Nuclear Reactor Regulation, NRC Bruce S. Mallett, Administrator, Region IV, NRC Mohan C. Thadani, Project Manager, NRC Resident Inspectors, CPSES
Attachment I to TXX-07013 Page 1 of I Inspection and Mitigation Summary for Comanche Peak Unit 1 Alloy 82/182 Pressurizer Butt Welds MRP-139 Volumetric Mitigation to be Nozzle Inspection Requirement Completed to be Met Comments Function Material Outage Start Date Outage Designation Description Designation (MM/YYYY) Designation Spray Post Structural Weld Overlay (SWOL) 1-PCV-0455B/C Alloy82/182 1RF12 02/2007 1RF12 Inspection Surge Alloy 82/182 1RF12 02/2007 IRF12 Post SWOL Inspection Safety 1-8010A Alloy 82/182 IRF12 02/2007 IRF12 Post SWOL Inspection Safety 1-8010B Alloy 82/182 IRF12 02/2007 IRF12 Post SWOL Inspection Safety 1-8010C Alloy 82/182 IRF12 02/2007 IRF12 Post SWOL Inspection Relief 1-8000A/B Alloy 82/182 02/2007. 1RF12 IRF12 Post SWOL Inspection to TXX-07013 Page 1 of 1 Inspection and Mitigation Summary for Comanche Peak Unit 2 Alloy 82/182 Pressurizer Butt Welds MRP-139 Volumetric Mitigation to be Nozzle Inspection Requirement Completed to be Met dComments Function Material Outage Start Date Outage Designation Description Designation (MM/YYYY) Designation Spray Alloy 82/182 2RFI0 03/2008 2RF10 Post Structural Weld Overlay (SWOL) 2-PCV-0455B/C Inspection Surge Alloy 82/182 2RF10 03/2008 2RF10 Post SWOL Inspection Safety 2-8010A Alloy 82/182 2RF10 03/2008 2RF10 Post SWOL Inspection Safety 2-8010B Alloy 82/182 2RF1O 03/2008 2RFIO Post SWOL Inspection Safety 2-8010C Alloy 82/182 2RF10 03/2008 2RFIO Post SWOL Inspection Relief 2-8000A/B Alloy 82/182 2RFIO 03/2008 2RFIO Post SWOL Inspection to TXX-07013 Page 1 of 23 JUSTIFICATION FOR PERFORMING UNIT 2 MITIGATION AFTER DECEMBER 31, 2007 A. Previous Inspection Results B. Assessment of Original Fabrication Weld Documentation C. Water Chemistry D. Plant Age E. Reactor Coolant System Leakage Monitoring F. Design and Residual Weld Stress Considerations.
F. 1 CPSES Transient and Fatigue Cycle Monitoring Program F.2 Effects of Residual Stress on ASME Code Design Qualification G. Overall Conclusions to TXX-07013 Page 2 of 23 A. PREVIOUS INSPECTION RESULTS ISI Exams:
The previous NDE examinations for the six Unit 2 Pressurizer nozzle to safe-end welds, having Alloy 82/182 weld material, occurred during the first interval. These examinations were all completed prior to the industry issues associated with the Alloy 600/82/182 material. These examinations followed the requirements of ASME Code Section XI, 1986 Edition, and consisted of volumetric (ultrasonic) and surface (penetrant) examination for each weld. The following table provides the examination information for each weld.
Previous Ultrasonic Examinations of Unit 2 Alloy 82/182 Pressurizer Butt Welds NOZZLE NON-PDI UT Examination Function/ Susceptible Material Outage Date Coverage Limitations Findings Designation Description Designation Surge Nozzle buttering-Alloy 2RF03 11/97 100% (1) None None 182 - Post Weld Heat Treatment (PWHT)
Weld - Alloy 82/182 Spray Nozzle buttering-Alloy 2RF02 3/96 100% (2) None None 182 - PWHT Weld - Alloy 82/182 Safety A Nozzle buttering-Alloy 2RF02 3/96 100% (3) None None 182 - PWHT Weld - Alloy 82/182 Safety B Nozzle buttering-Alloy 2RF02 3/96 100% (4) None None 182 - PWHT Weld - Alloy 82/182 Safety C Nozzle buttering-Alloy 2RF02 3/96 100% (5) None None 182 - PWHT Weld - Alloy 82/182 Relief Nozzle buttering-Alloy 2RF02 3/96 100% (6) None None 182 - PWHT Weld - Alloy 82/182 NOTES:
(1) Surge UT Exams - axial and circumferential scans - 450 shear axial 70 0 Longitudinal (L) axial 45 "L (2) Spray UT Exams - axial and circumferential scans - 45' L axial - 600 L (3) Safety A UT Exams - axial and circumferential scans - 45' L axial - 60' L (1 direction) axial - 300 shear (1 direction) to TXX-07013 Page 3 of 23 (4) Safety B UT Exams - axial and circumferential scans - 45' L axial - 10" L (I direction) axial - 300 shear (1 direction)
(5) Safety C UT Exams - axial and circumferential scans - 45" L axial - 10' L (1 direction) axial - 30' shear (1 direction)
(6) Relief UT Exams - axial and circumferential scans - 450 L axial - 60' L (1 direction) axial - 30" shear (I direction)
References:
- 3. Procedure TX-ISI-214, revision 0, "Ultrasonic Examination Procedure for Welds in Piping Systems and Vessels"
- 4. ASME Section XI, 1986 Edition Bare Metal Visual Inspections:
Bare metal visual inspections have been performed on all nozzle to safe end welds containing Alloy 82/182 weld material on the pressurizer. These inspections were performed starting in refueling outage 2RF07 (Fall 2003) and repeated in the next two refueling outages 2RF08 (Spring 2005) and 2RF09 (Fall 2006). 100% coverage was achieved on all locations with no indications of leakage. CPSES made the following commitments in response to NRC Bulletin 2004-01 regarding inspection requirements.
CPSES Commitment No. 27318 Each identified location in the CPSES Unit 1 and 2 pressurizers with Alloy 82/182 weld materials will be bare metal visual inspected each refueling outage, including 100% of the circumference over the axial length of the welds, until effective PWSCC-mitigative actions are taken or a technically robust, industry recommended inspection regime is issued. It is anticipated that direct visual examination will be performed. In areas where direct visual examination is not feasible or where remote techniques will result in equivalent examinations with reduced dose received, remote visual examination equipment may be used to perform the examination. Personnel performing the examinations will be qualified per the requirements for personnel implementing the CPSES Boric Acid Corrosion Detection and Evaluation Program per station procedure STA-737.
to TXX-07013 Page 4 of 23 Any accumulations of boric acid residue on or around the weld areas will be investigated to determine the origin of the deposit. If through wall leakage is suspected or if through wall leakage would be masked by leakage from other components, additional NDE techniques such as ultrasonic, eddy current or radiographic techniques will be used to characterize any indications. Should additional NDE techniques be utilized for follow up examinations, personnel involved will be qualified in accordance with ASME Section X1, 1989 Edition or later approved Code editions, if there are qualification requirements applicable to the examination technique(s) employed. ASME Code requirements for evaluation and repair of any flaws detected will be followed. These additional follow-up inspections will be documented on examination data sheets. (This commitment is contained within TXX-04140)
CPSES Commitment No. 27319 TXU Power will provide the requested information within 60 days after plant restart following the next inspection of the Alloy 82/182 pressurizer penetrations and steam space piping connections for CPSES Units I and 2. (This commitment is contained within TXX-04 140)
CPSES Commitment No. 27334 TXU Power will notify the NRC upon defining a flaw as circumferential cracking in an alloy 82/182 reactor coolant system piping attachment weld. (This commitment is contained within TXX-05056) to TXX-07013 Page 5 of 23 B. ASSESSMENT OF ORIGINAL FABRICATION WELD DOCUMENTATION Pressurizer Unit 2 Nozzle Safe End Component Fabrication Defect Identification and Repairs The following represents defects identified and repaired in the Unit 2 pressurizer nozzle safe end installations performed by the vendor. This information is based on Westinghouse report number PWROG PA-MSC-0233 dated September 11, 2006 and Receipt Inspection Report (RIR) 10317 dated May 15, 1979.
- 1) Pressurizer Surge Nozzle A. Indication found in surge nozzle weld buildup (for safe end).
Indication was removed and repaired using Alloy 182 prior to PWHT.
B. Linear indications identified in surge nozzle to safe-end weld.
Indications were removed via grinding, and repaired using Alloy 182.
During the grinding process nozzle base material was exposed and gouged. The base metal was weld repaired.
C. Following radiography testing an indication was identified in the pressurizer surge nozzle to safe-end weld location. The indication was removed and repaired with Alloy 182.
D. A defect noted in the previous weld repair was not fully repaired and an indication remained. Indication was removed, via grinding and repaired using Alloy 182.
- 2) Pressurizer Spray Nozzle A. The liner had 5 circumferential grooves, produced during the rolling process. The grooves were polished out.
- 3) Pressurizer Safety Nozzles A. Safety Nozzle A: Following connection of the safe-end and PWHT a ground out area on the bond line between the base metal of the nozzle forging and the nozzle build up was identified. The gouge was blended to a taper and inspected to ensure component remained within tolerances.
B. Safety Nozzle B had no reported defects or repairs.
C. Safety Nozzle C had no reported defects or repairs.
to TXX-07013 Page 6 of 23
- 4) Pressurizer Relief Nozzle A. An indication was identified on the cladding above the upper head to nozzle weld. The indication was removed, via grinding and the cladding was repaired using 308L.
B. Following PWHT an indication was identified in the nozzle inside diameter to upper head. It was determined to use as is since the PWHT process was completed. The defect did not violate the minimum wall thickness and no weld repair was required. Final ASME Code required non-destructive examinations were performed with acceptable results.
C. The safe-end to nozzle weld was rejected due to a defect.
The defect was removed via grinding and the weld was repaired using Alloy 182.
Field Installation Weld Records Records indicate that there were no weld repairs made to the six stainless steel safe-end to piping welds.
to TXX-07013 Page 7 of 23 C. WATER CHEMISTRY
Background
Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600 and its weldments continues to adversely affect Nuclear Power Plants that contain components with this alloy. Recently the industry has experienced cracking of austenitic stainless steel materials both foreign and domestically. Reactor Coolant System (RCS) components, welds, and penetrations contain these materials. The following discussion describes the CPSES Chemistry program regarding PWSCC mitigation.
Fluoride, Chloride, and Sulfate Control Anionic chemical species have been implicated in initiating and propagating classic Stress Corrosion Cracking (SCC) in austenitic stainless steels. Numerous industry studies have shown that fluorides, chlorides, sulfates, and dissolved oxygen are detrimental to these materials when coupled with the high stress and elevated temperature normally encountered within the RCS. While it is understood that PWSCC in Alloy 600 components can apparently occur in the absence of an aggravating chemical species, CPSES aggressively pursues minimizing these contaminants from the process water within the RCS to reduce the possibility of PWSCC initiation and propagation, in addition to classic SCC.
CPSES has maintained concentrations of RCS fluoride, chloride, and sulfate as low as reasonably achievable since commercial operation began for each unit. Data was specifically reviewed for this evaluation from the last two years, but this recent data is typical of that over the operating life of the units at CPSES. The review of the data demonstrated that these anions have been maintained well below the specifications and optimized limits. Specifications for these anions in the EPRI Pressurized Water Reactor Primary Water Guidelines, Revision 5 are set at 150 ppb (Reference C. 1). CPSES has established optimized limits of 50 ppb for each anion (Reference C.2). The following data are averages for each anion for both units over a two year period beginning 01/01/05 and ending 01/17/07: RCS average fluoride concentration is <3 ppb; RCS average chloride concentration is <4 ppb; and RCS average sulfate concentration is <2 ppb. No out-of-specification conditions or action levels have occurred for any of these anions during the two year period.
Dissolved Oxygen Control Because dissolved oxygen is one of the main factors in initiation of SCC, minimizing dissolved oxygen ingress to the RCS is imperative. Minimization of dissolved oxygen is also potentially important for mitigation of PWSCC. CPSES has stringent dissolved oxygen controls for two reactor coolant makeup sources: the reactor makeup water storage tanks (RMWSTs) and the boric acid storage tanks (BAST). The RMWSTs provide demineralized and deoxygenated makeup water and the BASTs provide boric acid solution (>7000 ppm boron) to the RCS.
to TXX-07013 Page 8 of 23 A bladder exists in each of the two RMWSTs to minimize dissolved oxygen intrusion from the atmosphere and a nitrogen cover gas is maintained above the bladders. Also, a nitrogen sparge is maintained when needed to minimize dissolved oxygen in the makeup water. RMWST dissolved oxygen concentrations typically are maintained at <20 ppb using these control measures.
The Boric Acid Storage Tank (two common tanks, one for each unit), is primarily used during scheduled shutdowns prior to refueling outages. The concentrated boric acid solution contained within this tank is used following reactor trip to rapidly borate the RCS. Since this tank is vented to the atmosphere and can contain dissolved oxygen concentrations approaching 8 ppm, it can be a significant source of oxygen intrusion into the RCS during shutdown while borating the RCS. A sparge ring has been placed in each tank so nitrogen can be sparged through the boric acid solution to remove dissolved oxygen prior to scheduled refueling outages.
Dissolved Oxygen Monitoring Coupled with the dissolved oxygen controls described on the previous page, CPSES has installed an in-line dissolved oxygen analyzer on each Chemical and Volume Control System (CVCS) inlet to the CVCS demineralizer sample line to continuously monitor RCS dissolved oxygen concentrations. The data generated filom this analyzer is compiled, archived and displayed by the Plant Computer System. The analyzer provides a rapid source of data to closely monitor RCS dissolved oxygen and is especially useful during scheduled shutdowns. Historically, RCS dissolved oxygen concentrations have remained well below specifications and optimized limits.
RCS Dissolved Hydrogen - Maintained High in the Band An EPRI study of the effects of hydrogen, pH, lithium and boron on primary water stress corrosion crack initiation in Alloy 600 describes a favorable impact to PWSCC by increasing the RCS dissolved hydrogen concentration to the upper portion of the band set by the EPRI primary guidelines (Reference C.3). The report shows that susceptibility to PWSCC is highest when the dissolved hydrogen concentration is approximately 29 cc/kg.
The model based on the data estimates a 16% increase in the Weibell Characteristic life as the concentration is increased from 29 cc/kg to 40 cc/kg. The graph generated from this model shows a general U shape to the curve with 29 cc/kg being the shortest time to crack initiation and longer initiation times arising as dissolved hydrogen decreases and increases from this point. Based on this data, CPSES has increased the RCS normal operation dissolved hydrogen concentration to >40 cc/kg. This concentration is sustained throughout the operating cycle. The EPRI Pressurized Water Reactor Primary Water Chemistry Guidelines recommend an RCS dissolved hydrogen range of 25 to 50 cc/kg for normal operation. The RCS dissolved hydrogen average for both units is approximately 43 cc/kg. Unit 2 has been operating in the higher portion of the band since May of 2005.
to TXX-07013 Page 9 of 23 Dissolved Hydrogen Monitoring CPSES has an in-line dissolved hydrogen analyzer on each Chemical and Volume Control System (CVCS) inlet to the CVCS demineralizer sample line to continuously monitor RCS dissolved hydrogen concentrations. The data generated from this analyzer is compiled, archived and displayed by the Plant Computer System. The analyzer provides a rapid source of data to closely monitor RCS dissolved hydrogen and is especially useful during scheduled shutdowns. Historically, RCS dissolved hydrogen concentrations have typically remained within the range recommended by EPRI, and based on the discussion in the previous paragraph, is maintained at >40 cc/kg.
Elevated pH(t)
The majority of the industry is operating the RCS at a pH above 6.9 pH (t) in accordance with the recommendations of the EPRI Primary Water Chemistry Guidelines for mitigation of corrosion product transport, fuel deposition, and radiation fields (Reference C. 1). The pH(t) is a calculated number, calculated using the EPRI Chem Works TM ,
Primary System pH Calculator software by using RCS boron, lithium, and ammonia concentration data and an RCS average temperature of 589°F. Comanche Peak has also operated at pH(t) 7.2 in the past, until the pH(t) 7.4 program was implemented for improved mitigation of corrosion product transport, fuel deposition, and radiation fields (Reference C. 1). This program was implemented for Comanche Peak Unit 2 in a phased approach. Cycle 7 was operated at a pH(t) of 7.3 after Cycle 6 operation of pH(t) of 7.2.
Cycles 8 and 9 were operated at pH(t) of 7.4, as well as the present cycle 10.
In addition to the benefits for which the program was implemented, the elevated pH program has been indicated to reduce susceptibility to PWSCC. (References C.3 and C.4). Evaluation of laboratory test data by EPRI of the effect of pH on PWSCC initiation concluded that raising the pH above 7.2 reduces the susceptibility to PWSCC. The EPRI evaluation concluded that the Weibull Characteristic life to PWSCC increased by 62%
(lower susceptibility to PWSCC initiation) when the pH at 589°F was increased fi-om 7.25 to 7.5 at 5890F (Reference C.3). Testing by Westinghouse also demonstrated that the Weibull Characteristic life to PWSCC increases by 45% (lower susceptibility to PWSCC initiation) when the pH is increased from 7.2 to 7.4 (Reference C.4). Given these multiple benefits, CPSES has elected to operate both units with the elevated pH(t) program.
Zinc Injection Studies have indicated that low concentrations of zinc injected into the reactor coolant can lower out of core radiation fields and also may potentially mitigate PWSCC (Reference C.5). CPSES has reviewed the current literature related to zinc injection in regard to the impact on the station's material perfonnance and reliability and has determined that at this time zinc injection is not considered to be appropriate for the CPSES primary chemistry program. Currently, CPSES is operating with the elevated pH(t) 7.4 program which may provide the same benefits as Zinc injection, without the to TXX-07013 Page 10 of 23 potential for Zinc effect on fuel crud deposits. Zinc is known to densify crud deposits in BWR fuel and has been implicated in the occurrence of axial offset anomaly in some industry experience with high duty cores. CPSES will continue to evaluate additional zinc injection data as it is published from demonstration plants with high duty cores. The zinc experience data will be compared to the data acquired after Unit 2 has completed the "Constant Elevated pH Demonstration" and the zinc injection program for CPSES will be reevaluated at that time.
Conclusion The primary system chemistry control program at Comanche Peak is designed and operated to minimize material degradation, such as primary stress corrosion cracking (SCC). One of the measures employed is the minimization of impurities that contribute to SCC. The levels of dissolved oxygen fluoride, chloride, and sulfate have been consistently well below industry action levels. Additional measures are taken with regard to makeup water as these tanks are treated with nitrogen to reduce dissolved oxygen.
Comanche Peak also utilizes dissolved hydrogen levels to control electrochemical potential and has implemented an elevated pH(t) program in the reactor coolant system.
Both of these measures have been demonstrated in EPRI research to lower susceptibility to SCC.
References:
C. 1 Pressurized Water Reactor Primary Water Chemistry Guidelines: Volume 1, Revision 5, EPRI, Palo Alto, CA: 2003. 1002884.
C.2 CPSES Primary Chemistry Strategic Plan, Revision 5 C.3 Material Reliability Program: Effects of Hydrogen, pH, Lithium and Boron on Primary Water Stress Corrosion Crack Initiation in Alloy 600 in the Range 320 - 330'C (MRP-147), EPRI, Palo Alto, CA: 2005.
1012145.
C.4 R. Jacko and R. Gold, "Crack Initiation in Alloy 600 Tubing in Elevated pH PWR Primary Water," 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, August 14-18, 2005; Salt Lake City, Utah C.5 Primary Water Reactor Primary Water Zinc Application Guidelines, EPRI, Palo Alto, CA: 2006. 1013420.
Attachment 3 to TXX-07013 Page 11 of 23 D. PLANT AGE The effect of plant aging and susceptibility to primary water stress corrosion cracking (PWSCC) of Alloy 600/82/182 is largely a function of time at temperature when all other variables are constant. Due to the high temperature, the Pressurizer is a susceptible location to PWSCC in an operating plant. Since the Pressurizer in a PWR operates at saturated conditions, all PWRs that operate at 2250 psi have a Pressurizer operating temperature within a few degrees of 653Y F, and can therefore be compared directly.
Comanche Peak Unit 2 is a relatively young plant compared to other US PWRs. EPRI MRP prepared a response to the NRC Bulletin 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles" on PWSCC. This document lists hours of operation for each of the 69 operating PWRs (MRP-48, Table 2-1). The EFPY data from the MRP-48 survey is shown below.
Comparison of EFPY of US PWRs Reported in MRP-48 EFPYs EFPYs EFPYs Rank Unit Name as of Rank Unit Name as of Rank Unit Name as of 2/2001 2/2001 2/2001 1 Ginna 23.9 24 North Anna 2 16.7 47 Waterford 3 12.4 2 Point Beach 1 22.9 25 Farley 2 16.4 48 Sequoyah 2 12.1 3 Point Beach 2 22.5 26 Cook 1 16.0 49 Byron 1 12.0 4 Prairie Island 1 22.4 27 ANO 2 15.9 50 Vogtle 1 11.9 5 Prairie Island 2 22.3 28 Palisades 15.6 51 Sequoyah 1 11.9 6 Kewaunee 21.6 29 Beaver Valley I 15.2 52 Catawba 2 11.7 7 Robinson 2 20.6 30 Crystal River 3 14.9 53 - Shearon Harris 11.6 8 Oconee 1 20.4 31 Davis-Besse 14.7 54 Byron 2 11.3 9 Oconee 2 20.3 32 St. Lucie 2 14.7 55 Palo Verde 1 11.1 10 Oconee 3 20.1 33 Millstone 2 14.0 56 Palo Verde 2 11.0 11 Fort Calhoun 19.9 34 Summer 13.9 57 Salem 2 10.8 12 SuIry 1 19.5 35 Callaway 13.8 58 Palo Verde 3 10.7 13 Surry 2 19.4 36 Indian Point 3 13.6 59 Vogtle 2 10.4 14 Turkey Point 3 19.3 37 McGuire 1 13.6 60 Braidwood 2 10.3 15 Turkey Point 4 19.0 38 San Onofre 2 13.5 61 Beaver Valley 2 10.2 16 St. Lucie 1 18.8 39 McGuire 2 13.4 62 Braidwood 1 9.9 17 Calvert Cliffs 1 18.3 40 San Onofre 3 13.3 63 Millstone 3 9.3 18 Farley 1 18.2 41 Cook2 13.3 64 South Texas 1 9.2 19 ANO 1 18.0 42 Salem 1 13.1 65 South Texas 2 8.9 20 Calvert Cliffs 2 17.9 43 Diablo Canyon 1 13.1 66 Comanche Peak 1 8.9 21 North Anna 1 17.1 44 Diablo Canyon 2 12.8 67 Seabrook 8.6 22 Indian Point 2 16.9 45 Wolf Creek 12.7 68 Comanche Peak 2 6.4 23 TMI 1 16.8 46 Catawba 1 12.5 69 Watts Bar I 4.3 to TXX-07013 Page 12 of 23 The list was current as of February 28, 2001. The relative position of Comanche Peak Unit 2 does not change as all the plants have increased operational hours since the table was compiled.
CPSES Unit 2 has accumulated 11.8 EFPY as of December 31, 2006.
to TXX-07013 Page 13 of 23 E. REACTOR COOLANT SYSTEM LEAKAGE MONITORING Comanche Peak uses a RCS Leakage monitoring program as described in station procedure, OPT-303, "Reactor Coolant System Water Inventory." Data from the Plant Computer is entered into a computer program that calculates the RCS leak rate in gallons per minute. This is required by Technical Specifications to occur every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with exceptions during periods when parameters are not stable. The calculated leak rate is broadcast on the morning Plant Status message each day and is discussed at the Plan of the Day meeting. A yearly trend for the leakage calculations is available to the entire site through the site intranet.
If leakage is greater than 0. 15 gpm sustained leakage (two or more OPT-303 calculations in a row) or 0.30 gpm calculated (single sample) leakage, Operations is directed to investigate the source of increased leakage. This requires documentation in the site Corrective Action Program. At 0.5 gpm sustained leakage rate, the procedure requires station management to consider an orderly shutdown. This leak rate is significantly less than the Technical Specification requirement of 1.0 gpm unidentified leakage. As an example of the conservative decision making process employed by station management, on February 3, 2005 management directed the shutdown of Unit I to repair a 0.66 gpm secondary system leak in containment.
The current values for Unit 2, based on one year worth of data, are:
Mean 0.0133 gpm 3 Standard Deviations 0.0554 gpm Sum 0.0687 gpm This would statistically indicate that the Comanche Peak Unit 2 baseline leak rate is approximately 0.0 133 gpm and that 99.9% of the data would fall below 0.0687 gpm.
This data includes one occurrence of a leak that caused the value to be greater than 0.1 gpm for the one calculation until the leak could be resolved.
Identifying small sources of leakage at an early stage ensures that new leakage sources are not masked. For example, currently on Unit 1, RCS Leak Rate calculations are being performed daily due to a leak of 0.008 gal/min.
Other methods for monitoring RCS leakage include:
Containment Sump Level and Flow, Containment Radiation, Containment Temperature, Containment Humidity, Containment Cooler Condensate Flow, Pressurizer Level and Pressure, Volume Control Tank Level, Automatic Makeup System (to the Chemical and Volume Control System),
to TXX-07013 Page 14 of 23 Safeguards Building Sump Level.
Alarn Procedures (ALM) exist for abnormal conditions in each of the above indications which direct the operator to enter the appropriate Abnormal Operating Procedure (ABN) for the plant condition and perform leak rate calculations as necessary. Unexpected changes in these parameters would necessitate documentation in the Corrective Action Program and resolution of the source(s) of leakage.
Comanche Peak has been involved in the development of the PWR Owner's Group guideline documents for RCS leakage monitoring: "Standard Process and Methods for Calculating RCS Leak Rate for Pressurized Water Reactors" (WCAP- 16423-NP, Rev. 0) and "Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors"(WCAP-16465-NP, Rev.0). The site provided six months of data as a pilot plant during the development phase.
Conclusion CPSES RCS leakage monitoring is already enhanced from that required by the plant Technical Specifications. If leakage is greater than 0. 15 gpm sustained leakage (two or more OPT-303 calculations in a row) or 0.30 gpm calculated (single sample) leakage, Operations is directed to investigate the source of increased leakage. At 0.5 gpm sustained leakage, the procedure requires station management to consider an orderly shutdown.
to TXX-07013 Page 15 of 23 F. DESIGN AND RESIDUAL WELD STRESS CONSIDERATIONS Loadings for the Pressurizer Nozzles The piping loads of Comanche Peak Unit 2 are provided in the following Table I so that the relative likelihood of cracking can be evaluated. The loadings were determined in six components: three force directions and three moments. The most damaging loadings for circumferential flaws are the bending moments, and the moments acting to potentially open a circumferential flaw have been summarized and reported as "Applied Stress."
The maximum bending stress which would be allowed according to Section III of the ASME Boiler & Pressure Vessel Code is reported as "Allowable Stress". This allowable stress has been reduced from the stress which would cause the pipe to fail by the required Section III margin of 2.77. Thus, it is called an allowable stress rather than a critical stress. Results are provided for Alloy 182 weld, for all three Pressurizer nozzle types (Reference F. 1).
Table 1: Applied and Allowable Stress Comparison:
Comanche Peak Design Margin Applied Allowable Allowable/Applied Comments Stress Stress Weld Nozzle ab Pb (ksi) (ksi)
Alloy Surge 2.968 6.966 2.35 >1.0 182 Spray 4.064 8.895 2.19 > 1.0 Safety/Relief 3.311 9.237 2.79 >1.0 Assessment of Comanche Peak Unit 2 Pressurizer Fatigue Status Fatigue management of the Comanche Peak Unit 2 Pressurizer lower head and Pressurizer surge nozzle is accomplished using the Westinghouse software WESTEMSTM(see Section F. 1). This task utilizes automated data transfer between the Comanche Peak Site and the Westinghouse Waltz Mill Site.
WESTEMSTM models developed specifically for Comanche Peak simulate the combined thermal hydraulic system from the RCS loop piping connection through the Pressurizer surge line up into the Pressurizer vessel including the upper head and spray nozzle locations. The thermal hydraulic model accounts for thennal stratification in the Pressurizer surge line as well as insurge /outsurge thermal stratification effects of the Pressurizer vessel proper. The mechanical simulation addresses both moments due to thermal stratification of the surge line and normal thermal expansion of the surge line.
Westinghouse perforrs the daily surveillance of Comanche Peak Unit 2 utilizing the WESTEMSTM code (Reference F.2). The stress and fatigue evaluations performed by WESTEMS TM follow the methodology of ASME Boiler & Pressure Vessel Code, Section to TXX-07013 Page 16 of 23 III, Subsection NB-3200. The usage factor is updated daily and reported monthly.
Acceptance for fatigue compliance is a usage factor less than 1.0.
The original stress reports referenced the ASME Boiler and Pressure Vessel Code, 1971 Edition, through the Summer 1973 Addenda. While the evaluation follows the 1989 Code structure, there is no substantial difference in the basic requirements of NB-3200 between the versions; therefore, the evaluation was performed consistent with the Comanche Peak Units 1 and 2 Pressurizer design specifications.
The original design fatigue values for the Pressurizer locations of concern are shown in the table below. These values do not include the effects of surge line stratification, which was not known to exist at the time of the original design.
Cumulative Usage Factors for Original Design:
Location Design Fatigue Usage (40 yrs.)
Spray Nozzle 0.299 Safety & Relief Nozzles Exempted from fatigue Surge Nozzle 0.033 ID - 0.0025 OD Cumulative Usage Factors as of January 2007 (11.9 EFPY):
Nozzle Value Spray 0.111 Safety/Relief 0.0 Surge 0.046 Per Reference F.3 the Pressurizer lower head and surge nozzle were reanalyzed to address the thermal stratification effects of insurge events which increased the design fatigue number at the nozzle ID from 0.033 to 0.1391 for 40 years. The Pressurizer surge nozzle and lower head locations were added to the WESTEMSTM monitoring program in March of 2002 (see Section F.1 for additional information).
The features of WESTEMSTM that are used to perform the monitoring evaluations are as follows:
- Pressurizer Thermal Hydraulic Model
- Global-to-Local (GTL) transformation to determine local temperatures
- Forcing Functions Generation (which drive various mechanical interaction models)
- ASME Design Stress and Fatigue at specified locations The last reported cumulative usage factor for the Pressurizer surge nozzle safe end location as of June 2006 was 0.04579 (Reference F.4).
The current cumulative usage factor reported by the WESTEMSTM system as of January 16, 2007, for the Pressurizer surge nozzle safe end location is 0.04583.
to TXX-07013 Page 17 of 23 Residual Weld Stress Assessment of the above Categories Primary Water Stress Corrosion Cracking (PWSCC) is a mechanism that could challenge the structural integrity of Alloy 82/182 welds in the Pressurizer safe-end for the surge line, spray line, and relief lines. In these materials, the time to initiate a crack and to grow that crack is related to chronic tensile stress, and strongly dependent on temperature.
Since pressurizers for virtually all PWRs operate at about 653 degrees F (saturated steam at 2250 psia) the temperature factor creates little difference between plants.
Unrelieved residual stress (due to weld shrinkage) causes zones of tensile and compressive stress throughout the pipe wall. At the locations of interest, namely alloy 82/182 butt weld locations, it is possible that chronic residual stresses, could combine with applied loads (pressure, thermal, connecting piping loads) and result in tensile membrane stresses sufficient to initiate and propagate a crack at the fluid-exposed ID surface of the pipe.
Current Cumulative Usage Factors (CUFs) reported for CPSES Unit 2 based on actual plant data and analyzed using WESTEMS TM are negligible (<0.05). Assuming that CUFs continue to accumulate at the same rate, fatigue life expectancy for the CPSES Pressurizer welds would be expected to last for many years (0.139/.05 x 40 yrs = 111 yrs). Significant margin therefore exists from a fatigue standpoint to accommodate potential adverse impact from PWSCC.
The available fatigue operating margins as well as design margins (greater than 2.0) documented above are unaffected by residual stresses that could potentially be introduced from a fabrication or repair process; therefore given the unlikely risk from PWSCC to these low age joints the CPSES Pressurizer will continue to perform its intended design function. The basis for this conclusion is provided in detail in Section F.2 It should be noted that RCS water environment ability to reduce fatigue resistance of RCS piping components was a condition that was previously evaluated under NUREG/CR-5999 and no adverse effects were identified from that effort on the fatigue usage factor for selected bounding components that included pressurizer nozzles.
Conclusion The likelihood of a failure in an 82/182 butt weld material is higher with increasing induced levels of local or residual stress. The applied stress at CPSES has been confinned to be well below acceptable limits. Additionally, the Pressurizer is a well supported component and the design analysis accounting for piping interface loads has revealed that the Pressurizer interface loads are moderately low. Thus it is highly unlikely that catastrophic crack propagation would result from cyclic fatigue. Sufficient operating and design margin from a fatigue and cumulative cyclic loading standpoint exists to accommodate the controlled amount of residual stress that can be induced from a fabrication and repair process.
to TXX-07013 Page 18 of 23
References:
F.l. WNET-130 (TCX) Vol. 1, Rev. 3, "Model D Series 84 Pressurizer Stress Report for Texas Utilities Services, Inc. Comanche Peak Unit 2", Westinghouse NSD Pensacola Plant, Pensacola, FL, March 1992 F.2. WESTEMS version 4.4 Westinghouse Proprietary Software for Integrated Diagnostics and Monitoring (Westinghouse letter LTR-PAFM-05-49, "Software Release Letter for WESTEMS 4.4", 7/08/2005, and Westinghouse calculation note CN-PAFM-05-33, Revision 0, "WESTEMS 4.4 Validation".
F.3. WPT-16287, "TXU ENERGY COMANCHE PEAK STEAM ELECTRIC STATION UNITS I and 2 Analysis of Pressurizer Insurge/Outsurge Analysis",
March 4, 2002 F.4. CPSES Engineering Report ER-ME-103, Revision 7, June 21, 2006 to TXX-07013 Page 19 of 23 F.1 CPSES TRANSIENT AND FATIGUE CYCLE MONITORING PROGRAM In early 1990 the Comanche Peak Transient and Fatigue Cycle Monitoring Program (STA-706) used manual methods to record and capture data to identify fatigue significant cycles experienced by the plant equipment. In December 1991 Westinghouse had completed WCAP-1 3122 (Reference F. 1.1) "Transient and Fatigue Cycle Monitoring Program for Comanche Peak Steam Electric Station Unit 1" which outlined a comprehensive program that addressed component selection, equipment limits, and instruments to be considered. By 1995 advances in technology made it possible to create the first fully automated Thermal Event Monitoring System called TEMS. This system was capable of capturing and identifying fatigue significant events at a component specific level automatically (Reference F.1.2).
Comanche Peak continued to maintain a leadership role in adopting new technology in this area when, in 1999, the system was upgraded to utilize Westinghouse's latest technology-WESTEMSTM. This resulted in a dramatic improvement in the accuracy and scope of the transient and fatigue cycle monitoring program.
The scope of the monitoring was expanded to include all of the locations affected by the NRC Bulletin 88-08 addressing thermal stratification and cycling in unisolable lines.
This involved automating the surveillance activities defined in the technical procedure ECS-35.01-31, "Evaluation of Thermal Monitoring Data."
When thermal stratification of the Pressurizer lower head and shell became an issue (Reference F.1.3) and Westinghouse recommended performing plant specific evaluations and monitoring, Comanche Peak addressed these issues by expanding the scope of the program. The current Pressurizer model simulations employed for both units include all of the physics model necessary to predict thermal stratification states in both the Pressurizer surge line and vessel.
Currently Comanche Peak is using the latest version of WESTEMSTM which employs a combination of advance computational methods coupled with artificial intelligence techniques to achieve a high degree of accuracy in its simulations and evaluations. The advance methods include a programmable generic algorithm to solve complex systems using multiple physics models and logic simulations coupled to expert systems artificial intelligence engine that performs ASME Boiler and Pressure Vessel Code Section III qualification reports.
Comanche Peak's intention is to continue at the forefront of employing automated solutions to these complex surveillance issues. The current status of this program is as follows:
- Westinghouse receives plant data daily from both units.
- The data is processed at the Energy Center by Westinghouse Personnel.
- A report is issued annually summarizing the observations for the current monitoring period.
to TXX-07013 Page 20 of 23
- Notable events and status are communicated to the responsible Comanche Peak engineer.
The components in the current scope include:
- Reactor Vessel
- Primary Loop Piping
- Steam Generator Primary Sides
- Pressurizer Surge Line
- Pressurizer Vessel
- Selected NRC Bulletin 88-08 monitoring locations Component Summary:
- Overall transient cyclic activity remains well within design cyclic assumptions.
- Transient behavior with respect to major components (with the exception of the Pressurizer) remains well within the design assumptions.
- Pressurizer Lower head stress and fatigue monitoring shows that vessel insurge and outsurge effects are being managed very well with fatigue damage accumulation rates well below any alarm thresholds for an anticipated 60 years of operations.
NRC Bulletin 88-08 Surveillance Activities:
Automated temperature monitoring and evaluation of the following locations is ongoing:
- Residual Heat Removal Loop 1
- Residual Heat Removal Loop 4
- Safety Injection to Loop I
- Safety Injection to Loop 2
- Safety Injection to Loop 3
- Safety Injection to Loop 4
- Normal Charging Line
- Alternate Charging Line
- Pressurizer Auxiliary Spray Line
References:
F. 1.1 WCAP-13 122 "Transient and Fatigue Cycle Monitoring Program for Comanche Peak Steam Electric Station Unit 1: Component Selection, Equipment Limits and Instrumentation List," December 1991, Westinghouse Proprietary Class 2 F. 1.2 WCAP- 14250 "Thermal Event Monitoring: Technical Report and System Guide," November 1995 F. 1.3 WCAP-14950, "Mitigation and Evaluation of Pressurizer Insurge/Outsurge Transients," February 1998 to TXX-07013 Page 21 of 23 F.2 EFFECTS OF RESIDUAL STRESS ON ASME CODE DESIGN QUALIFICATIONS This technical note was developed to discuss the impact of residual stresses on ASME Boiler and Pressure Vessel Code design rules. In summary there is a strong technical basis for the absence of any significant impact.
There are two major criteria which need to be met to qualify the design of a nuclear component, the first having to do with the maximum allowable loads, and the second having to do with the effects of cyclic fatigue.
Allowable loads: The approach used for the code rules in this area is to determine the ductile limit load for the component, and then apply an appropriate safety factor, the result being the maximum allowable stress intensity. The stresses which affect the ductile limit load are the primary stresses, which results from piping loads and internal pressure.
Stresses such as thennal or residual stresses have no impact on the ductile limit load, and this has been demonstrated through multiple experimental results.
Fatigue evaluation: The determination of the fatigue life of a nuclear component is done using the Miner's rule method of fatigue damage assessment. The fatigue design curves for the materials of interest are found in ASME Boiler and Pressure Vessel Code,Section III, Appendix I. The curves are specific to classes of materials, and incorporate a factor of two on stress, and 20 on cycles, to ensure safe operation. In addition to these margins, the fatigue design curves have been adjusted for the effect of R Ratio, which is defined as the ratio of minimum stress to maximum stress for a transient. Higher R ratios have been found to reduce fatigue life, so they have been adjusted to account for the maximum R Ratio possible.
When a region of a component has residual stresses present, their impact on fatigue loadings is additive for both the minimum and maximum values of the stress range during any given cycle. The net result is an increase in the R Ratio for any region with residual stresses, but this does not affect the design life evaluation, because the design curve has already been corrected for the maximum R Ratio. For this reason, residual stresses do not affect fatigue life evaluations performed using the ASME Boiler and Pressure Vessel Code.
to TXX-07013 Page 22 of 23 G. OVERALL CONCLUSIONS:
The administrative and technical justifications are summarized herein to permit Comanche Peak Unit 2 to complete Pressurizer butt weld inspection and mitigation activities in Spring 2008:
- 1. CPSES RCS leakage monitoring is already enhanced from that required by the plant Technical Specifications. If leakage is greater than 0.15 gpm sustained leakage (two or more OPT-303 calculations in a row) or 0.30 gpm calculated (single sample) leakage, Operations is directed to investigate the source of increased leakage. At 0.5 gpm sustained leakage, the procedure requires station management to consider an orderly shutdown.
- 2. The susceptibility to PWSCC is a function of time, material, environment, and temperature. The operating age of CPSES Unit 2 ranks 68t" of 69 US PWRs, (11.8 EFPY as of 12/3 1/2006) with no leakages experienced to date at these locations.
- 3. A significant number of butt welds have already been inspected in the industry, and there is no evidence of PWSCC of Alloy 82/182 butt weld safety significant failures.
- 4. MRP-109, "Materials Reliability Program: Alloy 82/182 Pipe Butt Weld Safety Evaluation for US PWR Plant Designs: Westinghouse and CE Plant Designs,"
provided butt weld safety assessment and bounding analyses, concluding that there are no immediate safety concerns associated with PWSCC of alloy 600 butt welds.
- 5. MRP 2007-003, "Implications of Wolf Creek Pressurizer Butt Weld Indications Relative to Safety Assessment and Inspection Requirements," concluded that careful visual inspection and enhanced on-line leakage monitoring ensures an extremely low risk of rupture until scheduled mitigation are completed for plants including CPSES Unit 2.
- 6. The likelihood of a failure in an 82/182 butt weld material is higher with increasingly induced levels of local or residual stress. The applied stress in CPSES has been confirmed to be well below acceptable limits. Additionally, the Comanche Peak Unit 2 Pressurizer is a well supported component and the design analysis accounting for piping interface loads has revealed that the Pressurizer interface loads are moderately low which minimizes the possibility of catastrophic crack propagation from cyclic fatigue. Sufficient operating and design margin from fatigue and cumulative cyclic loading standpoint exists to accommodate the controlled amount of residual stress that can be induced from fabrication and repair processes.
to TXX-07013 Page 23 of 23
- 7. Bare metal visual exams have been performed on all nozzle to safe end welds containing Alloy 82/182 weld material on the Pressurizer during every Comanche Peak Unit 2 refueling outage since Fall 2003 with no indications of leakage. The last bare metal visual exam was performed in October 2006.
- 8. The primary system chemistry control program at Comanche Peak is designed and operated to minimize material degradation, such as primary stress corrosion cracking (SCC). One of the measures employed is the minimization of impurities that contribute to SCC. The levels of dissolved oxygen fluoride, chloride, and sulfate have been consistently well below industry action levels. Additional measures are taken with regard to makeup water as these tanks are treated with nitrogen to reduce dissolved oxygen. Comanche Peak also utilizes dissolved hydrogen levels to control electrochemical potential and have implemented an elevated pH(t) program in the reactor coolant system. Both of these measures have been demonstrated in EPRI research to lower susceptibility to SCC.
to TXX-07013 Page 1 of 1 Licensing* Basis Commitments for Units I and 2 Commitment Commitment Number Description 27419 Inspection and mitigation activities of Pressurizer Alloy 82/182 butt welds for CPSES Unit 1 are scheduled to be completed during the refueling outage in Spring 2007 and for CPSES Unit 2 are scheduled to be completed during the refueling outage in Spring 2008 27420 Future inspections of Pressurizer butt welds at Comanche Peak Units I and 2 will be performed in accordance with industry guidance (MRP-139).
The results of future inspections or mitigations of Pressurizer Alloy 82/182 butt weld locations will be reported to the NRC within 60 days of startup from the outage during which they were performned.
27421 CPSES will provide an update to the NRC as necessary regarding changes to our leakage monitoring program by March 31, 2007.
27422 Comanche Peak will evaluate the feasibility of plant modifications to install diverse leakage detection capability. This may include, but is not limited to, video monitoring of Pressurizer piping, acoustic monitoring in the area of the Pressurizer, sensitive humidity monitoring, and other methods currently under evaluation. Plans for any additional capability which reliably and meaningfully adds to our ability to diagnose primary system leakage, as well as installation schedules as appropriate will be submitted to the NRC by May 31, 2007.
27423 If Comanche Peak Units 1 or 2 should shut down prior to presently planned inspection/ mitigation outages due to unacceptable primary system leakage, and if the leakage cannot be confirmed to originate from a source other than the Pressurizer, a bare metal visual examination of Alloy 82/182 butt weld locations on the Pressurizer will be performed to determine whether the leakage originated at those locations.
NCTPR001.R01 TU ELECTRIC PAGE: 1 DATE: 01-30-07 TOTAL PLANT SYSTEM TIME: 08:13:20 COMMITMENT DATA FORM LAST UPDATE DATE: 01/30/2007 COMMITMENT TYPE REG NUMBER: 27419 STATUS: OPN COMMITMENT REGISTER NUMBER:
I. COMMITMENT DETAIL INFORMATION AUTHORITY : NRC PRIORITY : G PRIMARY ORG : ZTS UNITS : 1 X 2 LIC CONTACT : JCH DELETE CODE:
SUB: (II) INSERVICE INSPECTIONS (ASME) (1A) ONGOING COMMIT-TO STATUS INC
( ) ( )
( ) ( )
STATUS : OPN (IX) OPN (2X)
MILESTONE : REC (IX) REA (2X)
SCH'D COMPLETION DATE: 04/30/2007 (1X) 04/17/2008 (2X)
REQ'D COMPLETION DATE: (lX) (2X)
TITLE: INSPECTION AND MITIGATION ACTIVITIES OF PZR ALLOY 82/182 BUTT WELDS DESCRIPTION:
Inspection and mitigation activities of Pressurizer Alloy 82/182 butt welds for CPSES Unit 1 are scheduled to be completed during the refueling outage in Spring 2007 and for CPSES Unit 2 are scheduled to be completed during the refueling outage in Spring 2008.
COMMENTS:
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ALL COMMITMENT INFORMATION HAS BEEN REVIEWED TO BE COMPLETE AND ACCURATE AT THIS TIME EXCEPT AS MARKED.
/ /D/
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I. COMMITMENT DETAIL INFORMATION AUTHORITY : NRC PRIORITY : G PRIMARY ORG : ZTS UNITS :1 X 2 LIC CONTACT : DELETE CODE:
SUB: (1A)
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ONGOING COMMIT-TO STATUS INC (( ))
( ) ( )
STATUS : OPN (IX) OPN (2X)
MILESTONE : RED (lX) REB (2X)
SCH'D COMPLETION DATE: 10/01/2008 (lX) 10/01/2009 (2X)
REQ'D COMPLETION DATE: (lX) (2X)
TITLE: FUTURE INSPECTIONS PERFORMED IAW MRPI39/ REPORT TO NRC WITHIN 60 DAYS DESCRIPTION:
Future inspections of Pressurizer butt welds at Comanche Peak Units 1 & 2 will be performed in accordance with industry guidance (MRP-139).
The results of future inspections or mitigations of Pressurizer Alloy 82/182 butt weld locations will be reported to the NRC within 60 days of startup from the outage during which they were performed.
COMMENTS:
Commitment in effect after structural weld overlay completed for Unit 1 in IRF12 and Unit 2 in 2RF10.
II. COMMITMENT REFERENCE INFORMATION SOURCE
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REQ'D COMPLETION DATE: 03/31/2007 (1X) 03/31/2007 (2X)
TITLE: LEAKAGE MONITORING PROGRAM DESCRIPTION:
CPSES will provide an update to the NRC as necessary regarding changes to our leakage monitoring program by March 31, 2007.
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NCTPR001.R01 TU ELECTRIC PAGE: 1 DATE: 01-30-07 TOTAL PLANT SYSTEM TIME: 08:53:20 COMMITMENT DATA FORM LAST UPDATE DATE: 01/30/2007 COMMITMENT TYPE REG NUMBER: 27422 STATUS: OPN COMMITMENT REGISTER NUMBER I. COMMITMENT DETAIL INFORMATION AUTHORITY : NRC PRIORITY G PRIMARY ORG : ZTS UNITS N N 2 LIC CONTACT : JCH DELETE CODE:
SUB: (IB) ONE TIME ACTION-TO STATUS CLS ( )
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REQ'D COMPLETION DATE: (ix) 05/31/2007 (2X)
TITLE: DIVERSE LEAKAGE DETECTION CAPABILITY DESCRIPTION:
Comanche Peak will evaluate the feasibility of plant modifications to install diverse leakage detection capability. This may include, but is not limited to, video monitoring of Pressurizer piping, acoustic monitoring in the area of the Pressurizer, sensitive humidity monitoring, and other methods currently under evaluation. Plans for any additional capability which reliably and meaningfully adds to our ability to diagnose primary system leakage, as well as installation schedules as appropriate will be submitted to the NRC by May 31, 2007.
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NCTPR001.R01 TU ELECTRIC PAGE: 1 DATE: 01-30-07 TOTAL PLANT SYSTEM TIME: 08:53:30 COMMITMENT DATA FORM LAST UPDATE DATE: 01/30/2007 COMMITMENT TYPE REG NUMBER: 27423 STATUS: OPN COMMITMENT REGISTER NUMBER:
I. COMMITMENT DETAIL INFORMATION AUTHORITY : NRC PRIORITY :G PRIMARY ORG : ZTS UNITS :1 X 2 LIC CONTACT : JCH DELETE CODE SUB: (1A) ONGOING COMMIT-TO STATUS INC (II INSERVICE INSPECTIONS (ASME)
( ) )
( )
STATUS : OPN (lx) OPN (2X)
MILESTONE : M5 (lx) M5 (2X)
SCH'D COMPLETION DATE:
(lx) (2X)
REQ'D COMPLETION DATE:
(lx) (2X)
TITLE: BARE METAL VISUAL EXAMINATIONS OF PZR ALLOY 82/182 BUTT WELDS DESCRIPTION:
If Comanche Peak Units 1 or 2 should shut down prior to presently planned inspection/mitigation outages due to unacceptable primary system leakage, and if the leakage cannot be confirmed to originate from a source other than the Pressurizer, a bare metal visual examination of Alloy 82/182 butt weld locations on the Pressurizer will be performed to determine whether the leakage originated at those locations.
COMMENTS:
Applicable for forced outages prior to mitigation of pressurizer Alloy 82/182 butt welds in Unit 1 and Unit 2.
II. COMMITMENT REFERENCE INFORMATION SOURCE
REFERENCE:
TYPE: LTR NUM: TXX-07-000013 SRC LVL: 0 SEC: REV:
PG : ISSUE DATE: 01/30/2007 OPEN/INCORPORATING/CLOSING
REFERENCE:
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SEC: REV:
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NCTPR001.R01 TU ELECTRIC PAGE: 2 DATE: 01-30-07 TOTAL PLANT SYSTEM TIME: 08:53:30 COMMITMENT DATA FORM LAST UPDATE DATE: 01/30/2007 COMMITMENT TYPE : REG NUMBER: 27423 STATUS: OPN COMMITMENT REGISTER NUMBER:
II. COMMITMENT REFERENCE INFORMATION OTHER
REFERENCE:
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NCTPR001.R01 TU ELECTRIC PAGE: 3 DATE: 01-30-07 TOTAL PLANT SYSTEM TIME: 08:53:30 COMMITMENT DATA FORM LAST UPDATE DATE: 01/30/2007 COMMITMENT TYPE REG NUMBER: 27423 STATUS: OPN COMMITMENT REGISTER NUMBER ALL COMMITMENT INFORMATION HAS BEEN REVIEWED TO BE COMPLETE AND ACCURATE AT THIS TIME EXCEPT AS MARKED.
NAME (PLEASE PRINT)EXT.DE_/PARTMENT DAT NAME (PLEASE PRINT) EXT. DEPARTMENT SIGNATURE DATE
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NAME (PLEASE PRINT) EXT. DEPARTMENT DATE SIGNATURE
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FOR LICENSING COMPLIANCE GROUP ONLY TNL REVIEW: 'I_
DATE SIGNATURE CT DATA ENTRY: D/
DATE
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INITIALS
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