3F0103-03, Response to Request for Additional Information, Bulletin 2002-01, Reactor Pressure Vessel Degradation & Reactor Coolant Pressure Boundary Integrity

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Response to Request for Additional Information, Bulletin 2002-01, Reactor Pressure Vessel Degradation & Reactor Coolant Pressure Boundary Integrity
ML030440225
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 01/28/2003
From: Young D
Progress Energy Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0103-03, BL-02-001
Download: ML030440225 (18)
v • d • e


Text

Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 50.54(0 January 28, 2003 3F0103-03 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Request for Additional Information, Bulletin 2002 01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity"

References:

1. NRC to FPC letter, 3Nl102-06, dated November 22, 2002, Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity," 60-Day Response for Crystal River Unit 3 Request for Additional Information (TAC No. MB4539)
2. FPC to NRC letter, 3F0502-01, dated May 15, 2002, Crystal River Unit 3 Day Response to Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity"
3. FPC to NRC letter, 3F0302-1 1, dated March 28 2002, Crystal River Unit 3 Response to NRC Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity"

Dear Sir:

Reference 1 contains nine questions regarding the Crystal River Unit 3 (CR-3), 60-Day Response to Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity." Answers to those questions are provided in the Attachment to this letter.

The Attachment provides the basis for concluding that CR-3's Boric Acid Corrosion, Inspection and Evaluation Program is in compliance with the applicable regulatory requirements discussed in GL 88 05 and NRC Bulletin 2002-01. Additionally, the program incorporates plant and industry operating experience. The program will continue to be evaluated and enhanced, as needed, incorporating industry experience and best practices.

This letter establishes no new regulatory commitments.

Progress EnerW Florida, Inc.

Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428

U. S. Nuclear Regulatory Commission Page 2 of 3 3F0103-03 Although not required by CR-3 procedures, this Request for Additional Information response has been reviewed by the Plant Nuclear Safety Committee and comments were incorporated.

If you have any questions regarding this submittal, please contact Mr. Sid Powell, Supervisor, Licensing and Regulatory Programs at (352) 563-4883.

Sincerely, Dale E. Young Vice President, Crystal River Nuclear Plant DEY/lvc

Attachment:

Response to Request for Additional Information, Items 1 Through 9 Regarding Crystal River Unit 3, 60-Day Response for NRC Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation And Reactor Coolant Pressure Boundary Integrity" xc: NRR Project Manager Regional Administrator, Region HI Senior Resident Inspector

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT Response to Request for Additional Information, Items 1 Through 9 Regarding Crystal River Unit 3, 60-Day Response for NRC Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity"

U. S. Nuclear Regulatory Commission Page I of 15 3F0103-03 Attachment

==

Introduction:==

The NRC staff's review of the licensees' responses to Bulletin 2002-01 resulted in the following Request for Additional Information (RAI). In accordance with NRC's request, Progress Energy Florida, Inc. is providing the NRC questions and the responses for Crystal River Unit 3 (CR-3) to the RAI. The information provided below, in conjunction with information previously provided, constitute the basis for concluding that CR-3's boric acid inspection program is providing reasonable assurance of compliance with the applicable regulatory requirements discussed in Generic Letter 88-05 and Bulletin 2002-01.

Question 1:

Provide detailed information on, and the technical basis for, the inspection techniques, scope, extent of coverage, and frequency of inspections, personnel qualifications, and degree of insulation removal for examination of Alloy 600 pressure boundary material and dissimilar metal Alloy 82/182 welds and connections in the reactor coolant pressure boundary. Include specific discussion of inspection of locationswhere reactorcoolant leaks have the potential to come in contact with and degrade the subject material (e.g., reactor pressurevessel bottom head).

Response

The technical bases for examination of Alloy 600 pressure boundary material and dissimilar metal Alloy 82/182 welds and connections in the reactor coolant pressure boundary (RCPB) is consistent with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, and Generic Letter 88-05.

In addition to those Alloy 600 components to be examined under the rules of ASME Section XI, CR-3 has augmented the Inservice Inspection (ISI) Non-Destructive Examination (NDE) Program to include Reactor Pressure Vessel (RPV) penetrations, Control Rod Drive Mechanisms (CRDM) nozzles, and other Alloy 600 components. The Babcock & Wilcox Owners Group (BWOG) Materials Committee has performed Primary Water Stress Corrosion Cracking (PWSCC) susceptibility reviews. CR-3 has used that ranking as a basis for the Augmented Alloy 600 Program examinations. The augmented inspection encompasses the interfaces where reactor coolant leaks have the potential to come in contact and produce RCPB degradation. Augmented components are VT-2 examined for evidence of leakage, including boric acid residue, following insulation removal. The augmented component examinations are scheduled to coincide with the 10-year ASME XI ISI component examination schedules. ASME Code, and augmented examinations, are performed by ASNT-TC-1A qualified and certified examiners. Examiners performing augmented VT-2 examination receive additional training to recognize the characteristics of small volume boric acid leakage. CR-3 has completed approximately 50% of the augmented Alloy 600 weld examinations (bare metal) for this interval. The results of the visual inspection of the CRDM nozzle penetrations performed during Refueling Outage 12 (fall 2001) and the corrective actions taken as a result of leakage from a CRDM nozzle were provided in Florida Power Corporation (FPC) to NRC letter, 3F1 101-04, dated November 19, 2001, Crystal River Unit 3 - Information Requested in Item 5 of NRC Bulletin 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles."

U. S. Nuclear Regulatory Commission Page 2 of 15 3F0103-03 Attachment The following table identifies the susceptible Alloy 600 pressure boundary components and Alloy 82/182 welds at CR-3. Also included in the table are the inspection techniques, frequencies, degree of insulation removal and types of insulation.

Component (Alloy 600 Quantity Inspection Extent of Frequency Degree of Insulation pressure boundary material Techniques Coverage Insulation Type and Alloy 82/182 welds) Removal Core Flood Tank (CFT) Level 2/Tank Visual 100% 1 / 40 months Not insulated N/A Sensing Nozzles (VT-2) per Section XII Core Flood Tank (CFT) I/Tank VT-2 100% 1 / 40 months Not insulated N/A Make-up Nozzle per Section XI Core Flood Tank (CFT) 1/Tank VT-2 100% 1 / 40 months Not insulated N/A Outlet Weld per Section XI Core Flood Tank (CFT) I/Tank VT-2 100% 1 / 40 months Not insulated N/A Pressure Relief Nozzle per Section XI Core Flood Tank (CFT) 2/Tank VT-2 100% 1 /40 months Not insulated N/A Pressure Sensing Nozzle per Section XI Core Flood Tank (CFT) 1/Tank VT-2 100% 1 / 40 months Not insulated N/A Sample Connection per Section XI Once Through Steam I/OTSG VT-2 100% 1/ 120 months 100% Reflective Generator (OTSG) Primary per Augmented Drain Program Pressurizer Lower Level 3 VT-2 100% 1/ 120 months 100% Reflective Sensing Nozzle per Augmented Program Pressurizer Pressure Relief 3 Penetrant 100% 1/ 120 months 100% Fiberglass Nozzle weld Test per Section XI (PT)

Pressurizer Sample Nozzle 1 VT-2 100% 1 / 120 months 100% Reflective per Augmented Program Pressurizer Spray Nozzle Safe 1 VT-2 100% 1 / 120 months 100% Fiberglass End per Augmented Program essurizer Surge Nozzle 1 Ultrasonic 100% 1 / 120 months 100% Reflective eld Examination per Section XI

(_T_ /PT Pressurizer Thermowell 1 VT-2 100% 1 / 120 months 100% Reflective per Augmented Program Pressurizer Upper Level 3 VT-2 100% 1 / 120 months 100% Reflective Sensing Nozzle per Augmented Program Pressurizer Vent Nozzle 1 VT-2 100% 1 / 120 months 100% Fiberglass per Augmented Program Reactor Coolant System 1IT / PT 100% 1/ 120 months 100% Reflective (RCS) Decay Heat Nozzle per Section XI

U. S. Nuclear Regulatory Commission Page 3 of 15 3F0103-03 Attachment Component (Alloy 600 Quantity Inspection Extent of Frequency Degree of Insulation pressure boundary material Techniques Coverage Insulation Type and Alloy 82/182 welds) Removal Reactor Coolant System 3 VT-2 100% 1 / 120 months 100% Reflective (RCS) Drain Nozzle per Augmented Program Reactor Coolant System 1 VT-2 100% 1 / 120 months 100% Reflective (RCS) Drain Nozzle Safe End per Augmented Program Reactor Coolant System 4 VT-2 100% 1 / 120 months 100% Reflective (RCS) Flow Meter Nozzle per Augmented Program Reactor Coolant System 4 UT / PT 100% 1 / 120 months 100% Reflective (RCS) High Pressure per Section XI Injection Nozzle Weld I Reactor Coolant System 4 VT-2 100% 1 / 120 months 100% Reflective (RCS) Lower Cold Leg per Augmented Resistive Temperature Program Element Mounting Boss Reactor Coolant System 4 VT-2 100% 1 / 120 months 100% Reflective (RCS) Lower Cold Leg per Augmented Temperature Connections Program Reactor Coolant System 8 UT / PT 100% 1 / 120 months 100% Reflective (RCS) Piping Reactor Coolant per Section XI Pump Inlet / Outlet Welds Reactor Coolant System 1 UT / PT 100% 1 / 120 months 100% Reflective (RCS) Piping Surge Nozzle per Section XI Welds Reactor Coolant System 4 VT-2 100% 1 / 120 months 100% Reflective (RCS) Hot Leg Pressure Tap per Augmented Nozzle Program Reactor Coolant System 4 VT-2 100% 1 / 120 months 100% Reflective (RCS) Cold Leg Pressure Tap per Augmented Nozzle Program Reactor Coolant System 4 VT-2 100% 1 / 120 months 100% Reflective (RCS) Resistive Temperature per Augmented Element Mounting Boss Program Reactor Coolant System 2 VT-2 100% 1 / 120 months 100% Reflective (RCS) Temperature per Augmented Connection Program Reactor Coolant System 2 VT-2 100% 1 / 120 months 100% Reflective (RCS) Vent Nozzle per Augmented Program Reactor Vessel Core Flood 2 UT from 100% scan 1/ 120 months Not required N/A Weld Inside performed per Section XI Diameter (Actual (ID) calculated coverage is 86%)

U. S. Nuclear Regulatory Commission Page 4 of 15 3F0103-03 Attachment Component (Alloy 600 Quantity Inspection Extent of Frequency Degree of Insulation pressure boundary material Techniques Coverage Insulation Type and Alloy 82/182 welds) Removal Reactor Vessel Control Rod 2/Tube PT 6 motor 1 / 120 months 100% Reflective Drive Mechanism (CRDM) (136) tubes per Section XI Motor Tube Welds (2) examined Reactor Vessel Control Rod 1/Nozzle VT-2 100% Each Refueling 100% Reflective Drive Mechanism (CRDM) (69) Outage Nozzle to Head J-Groove Weld Reactor Vessel Control Rod 2/Nozzle VT-2 100% 1 / 120 months 100% Reflective Drive Mechanism (CRDM) (138) per Augmented Nozzle Welds (2) Program CRDM Nozzle Forgings I/Nozzle VT-2 100% 1 / 120 months 100% Reflective (69) per Augmented 1 Program Reactor Vessel Incore 52 VT-2 25% (See I / 120 months 33% Reflective Instrumentation Nozzles Response to per Section XI Question 2)

Reactor Vessel Monitor Tap 1 VT-2 100% 1 / 120 months 100% Reflective Weld per Augmented L_ I_ I I Program

U. S. Nuclear Regulatory Commission Page 5 of 15 3F0103-03 Attachment Question 2:

Provide the technical basisfor determining whether or not insulation is removed to examine all locations where conditions exist that could cause high concentrations of boric acid on pressure boundary surfaces or locations that are susceptible to primary water stress corrosioncracking (Alloy 600 base metal and dissimilarmetal Alloy 82/182 welds). Identify the type of insulationfor each component examined, as well as any limitationsto removal of insulation. Also include in your response actions involving removal of insulation required by your procedures to identify the source of leakage when relevant conditions (e.g., rust stains, boric acid stains, or boric aciddeposits) arefound.

Response

CR-3 removes all the insulation from the Alloy 600 and Alloy 82/182 welds to perform the examinations listed in the table in Question 1. The exception has been the bottom head Reactor Vessel Incore Instrumentation Nozzles. These nozzles have been determined to be low probability of failure per the B&W Owners Group (BWOG). Based on this, the examinations to date have consisted of examining the insulation while installed and general area reviews when the insulation is removed to perform scheduled ISI on the Reactor Vessel Support Skirt. VT-3 examination is performed on the interior of the reactor vessel support at three positions along the circumference of the support located 120 degrees apart.

Approximately 33% of the Reactor Vessel bottom insulation is removed to perform this exam. There has not been an observed accumulation of boric acid crystals in this area.

Based on recent Operating Experience (OE), CR-3 has scheduled the bottom head incore instrumentation nozzles for a complete VT-2 examination during the next refueling outage scheduled for fall 2003 (R13). The insulation will be removed to provide 100% access to the nozzles.

In addition, the CR-3 Boron Corrosion Control Procedure (PM-168) contains the following requirement when performing evaluations of components with observed boric acid deposits:

"NOTE: For any components where boron crystals prevent inspections of component parts a reinspection may be required after the boric acid has been removed from the component.

This is dependent upon the presence of known or suspected low-alloy or carbon steel parts.

If they are present, an inspection of the component a second time after removal of boric acid will be required to determine if degradation has occurred to these component parts."

U. S. Nuclear Regulatory Commission Page 6 of 15 3F0103-03 Attachment Question 3:

Describe the technical basisfor the extent andfrequency of walk-downs and the method for evaluating the potentialfor leakage in inaccessible areas. In addition, describe the degree of inaccessibility, and identify any leakage detection systems that are being used to detect potential leakagefrom components in inaccessibleareas.

Response

The walk-downs performed to detect evidence of leakage from borated systems are scheduled in a manner consistent with 10 CFR 50.65, 10 CFR 50.55 and CR-3 Improved Technical Specifications (ITS). Currently, the Boric Acid Corrosion Inspection, and Evaluation Program at CR-3 is integrated into other plant processes. The required inspections for this program are performed primarily by system engineers at the beginning of refueling outages and during plant shutdown (MODE 3). Plant walk-downs in the Reactor Building include the entire RCPB. Focused exams of specific areas (i.e., RPV Head, scheduled ISI) are performed during refueling outages when access is permitted. This methodology results in RCPB being completely accessible for examination.

These inspections are performed by a team of engineers who are briefed by the Boric Acid Corrosion Control program engineer on the requirements of GL 88-05 and PM-168. These engineers are trained to perform walk-downs.

The training addresses the following elements for the engineering staff: how to look for deficiencies (i.e., obvious and subtle indicators, symptoms, causes and consequences, understanding equipment function, identifying aggressive environment and other potential causes of degradation etc.), when to perform the observations (i.e., Maintenance Rule frequency for (a)(1) and (a)(2) systems, opportunistic, system being opened, etc.), where to look (i.e., everywhere) and what to look for (i.e., unusual sounds, wetness, trash, etc.). The EPT-359 training provided to system engineers is used to assure specific issues are identified. This training is also used to sensitize the system engineers on any recent industry events occurring at other nuclear sites (i.e., Oconee, TMI, VC Summer, etc.).

Classroom training is accomplished by EPT-359, "Engineers Role in Equipment Aging Management," which covers fundamentals of performing effective walk-downs. This allows engineers trained in performing walk-downs the opportunity to identify leakage, as well as other issues that may require work during the outage, early enough to get the work properly planned and added to the current outage schedule.

The training includes discussions of recent OE and requirements of PM-168 at the pre-job brief.

During each refueling outage, certified VT-2 inspectors perform visual exams on all the bolted connections on the Reactor Coolant System (RCS) as required by the ASME Code,Section XI. This exam is performed with the insulation removed. If boric acid residue is noted, the ASME code requires the bolting be removed and a VT-3 exam be performed on the bolting. Additionally, the source of the leakage is located and corrected, and the surrounding area is assessed in accordance with PM-168 (Response to Question 4). The

U. S. Nuclear Regulatory Commission Page 7 of 15 3F0103-03 Attachment ASME Code-required RCS system leak test is performed during plant startup (MODE 3) by certified visual inspectors. The inspection boundary includes the entire RCS. The CR-3 Corrective Action Program (CAP) is used to document, track, investigate and correct adverse conditions. At CR-3, OE is controlled under Action Tracking, apart of the CAP.

The radiation protection personnel performing decontamination evolutions are aware of boric acid corrosion impact and have instructions to document, via the CAP, any signs of boric acid accumulation observed on external surfaces of components. Operators are also performing walk-downs to determine leakage amounts and cleanliness per the applicable surveillance procedures (SP-317, "RCS Water Inventory Balance," and SP-324, "Containment Inspection").

CR-3 does not have an installed local leakage detection system, however, very small amounts of leakage can be detected. CR-3 currently monitors, tracks, and trends RCS leakage to hundredth of a gallon per minute. As described in the CR-3 Final Safety Analysis Report (FSAR), Reactor Coolant Pressure Boundary integrity can be continuously monitored in the control room by the surveillance of variation from normal conditions for the following:

a. Reactor building sump level
b. Reactor building radioactivity levels
c. Condenser off-gas radioactivity levels (to detect steam generator tube leakage)
d. Decreasing makeup tank water level (indicating system leakage)

Gross leakage from the reactor coolant boundary will also be indicated by a decrease in pressurizer water level and rapid increase in the reactor building sump water level as described in FSAR Section 4.2.3.8.

Appropriate actions are taken to identify leakage sources. Once the source is identified, the CAP is used to determine the appropriate corrective actions.

U. S. Nuclear Regulatory Commission Page 8 of 15 3F0103-03 Attachment Question 4:

Describe the evaluations that would be conducted upon discovery of leakage from mechanical joints (e.g., bolted connections) to demonstrate that continued operation with the observed leakage is acceptable. Also describe the acceptance criteria that were establishedto make such a determination. Provide the technical basis used to establish the acceptancecriteria.In addition,

a. if observed leakage is determinedto be acceptablefor continuedoperation, describe what inspection/monitoringactions are taken to trend/evaluate changes in leakage, or
b. if observed leakage is not determinedto be acceptable,describe what corrective actions are taken to address the leakage.

Response

When evidence of leakage (i.e., boric acid crystals, water, etc.) is found, plant procedure PM-168 and the plant's CAP require an assessment of the condition. PM-168 provides instructions for determining the amount of wastage, if any, and the impact on adjacent components. If the amount of wastage cannot be determined without the removal of the crystals, then guidance is provided to have the area cleaned and reevaluated after cleaning.

The goal of the program is to have no leaks left in service, but if this is not achievable (for non-RCPB leaks), guidance is provided to assess the leak, assess the estimated corrosion rate, determine the impact on adjacent components and then document the evaluation to allow continued service. These evaluations are documented in the CAP. This approach was developed based on the guidance contained in EPRI Report, TR-1027485, "Boric Acid Corrosion Guidebook," and is contained in PM-168. The evaluation of the component will include any reinspection/monitoring requirements.

Steps from the boric acid corrosion control (BACC) program procedure include the following considerations or requirements:

EVALUATIONS When evaluating components due to boric acid concerns, consider the need for a Design Change or Operating Procedure change. The goal of the BACC Program is to eliminate the leak source to reduce future inspections and maintenance activities.

Evaluation of an active leak and a justification for continued operations (JCO) and/or startup from a shutdown with a borated system leak shall include the following:

  • Characterize the leakage and degradation.
  • Predict leakage and degradation until repair can be implemented.
  • Assess future degradation versus code requirements for full qualification of component.
  • Establish subsequent inspection requirements.
  • Determine most probable failure mechanism AND predict effects of this failure to plant operations.

U. S. Nuclear Regulatory Commission Page 9 of 15 3F0103-03 Attachment

  • IF the leak is excessive (i.e., large cleanup activity, high probability of excessive wastage, etc.),

THEN recommend immediate corrective action.

  • Evaluate leaks found during a shutdown to include the following:

Determine if the component can be repaired online.

All inaccessible components should be repaired/replaced prior to startup unless a JCO has been performed for the condition.

Initiate a work request (WR) to clean, repair and/or replace component(s) if required.

If leak is determined to be active, ensure a corrective action document has been initiated to stop the leak (i.e., adjust packing, initiate WR, initiate CAP Document, etc.).

Active leaks are tracked.

If leakage is not determined to be acceptable, the schedule for repair/replacement activities is established.

U. S. Nuclear Regulatory Commission Page 10 of 15 3F0103-03 Attachment Question 5:

Explain the capabilitiesof your program to detect the low levels of reactorcoolant pressure boundary leakage that may result from through-wall cracking in the bottom reactor pressure vessel head incore instrumentation nozzles. Low levels of leakage may call into question reliance on visual detection techniques or installed leakage detection instrumentation,but have the potentialfor causing boric acidcorrosion. The NRC has had a concern with the bottom reactor pressure vessel head incore instrumentation nozzles because of the high consequences associated with loss of integrity of the bottom head nozzles. Describe how your program would evaluate evidence of possible leakage in this instance. In addition, explain how your program addresses leakage that may impact components that are in the leak path.

Response

The Boric Acid Control Program at CR-3 will detect low levels of reactor coolant leaks that may result from through-wall cracking in the bottom reactor pressure vessel head incore instrumentation nozzles through visual observation. The incore nozzles have been partially observed when the insulation was removed to perform scheduled ISI on the RV support skirt. Although the PWSCC susceptibility of the Incore Monitoring Instrumentation (IMI) nozzles is believed to be low due to low operating temperature (See response to Question 2),

the entire area is currently scheduled for 100% bare metal inspection during the next refueling outage (R13). During R13, the insulation will be removed to provide 100% access to the penetrations and the area will be VT-2 examined. The results will be documented and evaluated per the BACC Program (BACC Program Evaluation requirements are described in the response to Question 4 above, these requirements include the establishment of subsequent inspection requirements) and the CAP. The BACC Program Evaluation requirements also include an assessment of components that are in the leak path, that may be impacted by the leakage.

U. S. Nuclear Regulatory Commission Page 11 of 15 3F0103-03 Attachment Question 6:

Explain the capabilitiesof yourprogram to detect the low levels of reactorcoolant pressure boundary leakage that may result from through-wall cracking in certain components and configurationsfor other small diameter nozzles. Low levels of leakage may call into question reliance on visual detection techniques or installed leakage detection instrumentation, but have the potential for causing boric acid corrosion. Describe how your program would evaluate evidence of possible leakage in this instance. In addition, explain how your program addresses leakage that may impact components that are in the leak path.

Response

As explained in the response to Question 4 above, when evidence of leakage (i.e., boric acid crystals, water, etc.) is found, plant procedure PM-168 and the plant's corrective action program require an assessment of the condition. PM-168 provides instructions for determining the amount of wastage, if any, and the impact on adjacent components. If the amount of wastage cannot be determined without the removal of the crystals, then guidance is provided to have the area cleaned and reevaluated after cleaning. The goal of the program is to have no leaks left in service, but if this is not achievable (for non-through-wall leaks),

guidance is provided to assess the leak, assess the estimated corrosion rate, determine the impact on adjacent components and then document the evaluation to allow continued service. These evaluations are documented in the CAP. This approach was developed based on the guidance contained in EPRI Report TR-1027485 "Boric Acid Corrosion Guidebook," and is contained in plant procedure PM-168. The evaluation of the component will include any reinspection/monitoring requirements.

U. S. Nuclear Regulatory Commission Page 12 of 15 3F0103 -03 Attachment Question 7:

Explain how any aspects of your program (e.g., insulation removal, inaccessible areas,low levels of leakage, evaluation of relevant conditions) make use of susceptibility models or consequence models.

Response

The inspection frequency of the Alloy 600/82/182 material components was based on the BWOG research contained in the Framatome Technologies Inc. (FTI) document 51 5003018-00 titled, "Program Plan for Alloy 600 PWSCC Life Cycle Management." This proprietary document uses susceptibility models and industry experiences to calculate the "Relative Time to Failure." CR-3 inspects the susceptible Alloy 600 pressure boundary components identified in the response to Question 1.

The inspections have been added to the ISI NDE Program as augmented exams and are performed coincident with the ASME Code required examinations. The examinations are performed by certified and qualified inspectors.

U. S. Nuclear Regulatory Commission Page 13 of 15 3F0103-03 Attachment Question 8:

Provide a summary of recommendationsmade by your reactor vendor on visual inspections of nozzles with Alloy 600/82/182 material,actionsyou have taken or plan to take regarding vendor recommendations,and the basisfor any recommendationsthat are not followed.

Response

In addition to the BWOG research referenced in the response to Question 7, CR-3 has received documentation from Framatome ANP (log 4-02 file 205/T4.4 PSC 4-02) which includes the following statements of recommendation for the lower head IMI nozzles:

"Under ideal circumstances, NDE (UT/eddy current examination (ECT)/PT) would be the suggested corrective action. However, given the current state of qualified NDE techniques and accessibility concerns, a bare metal visual examination of the IMI nozzle/lower RV head interface area at all B&W-fabricated 177-FA reactor vessels should be performed at the earliest opportunity. Other actions may also be appropriate to consider."

Based on this recommendation and current industry experiences, CR-3 is planning a detailed VT-2 bare metal examination of the RPV bottom head IMI nozzle interface area during R13 currently scheduled for the fall of 2003. These exams have been added to the augmented ISI program and have been identified to outage management as required outage work.

U. S. Nuclear Regulatory Commission Page 14 of 15 3F0103-03 Attachment Question 9:

Provide the basis for concluding that the inspections and evaluations described in your responses to the above questions comply with your plant Technical Specifications and Title 10 of the Code of FederalRegulations, Section 50.55(a), which incorporatesSection XI of the American Society of Mechanical Engineers (ASME) Code by reference. Specifically, address how your boric acid corrosion control program complies with ASME Section XI, paragraphIWA-5250 (b) on correctiveactions. Include a description of the procedures used to implement the corrective actions.

Response

ASME Class I components (which include RCPB, Reactor Vessel Head (RVH) and Control Rod Drive Mechanism (CRDM) nozzles) must meet the requirements of Section XI of the ASME Boiler and Pressure Vessel Code. Table IWB-2500-1 of Section XI provides examination requirements for welds and references IWB-3000 for acceptance standards.

IWA-5250 (b) states, "If boric acid residues are detected on components, the leakage source and the areas of general corrosion shall be located. Components with local areas of general corrosion that reduce the wall thickness by more than 10% shall be evaluated to determine whether the component may be acceptable for continued service, or whether repair/replacement activities will be performed." While this comes from a later edition of Section XI than the CR-3 ASME Code of record, CR-3 program essentially have the same requirements to locate leakage and assess the impact of any corrosion on the structural integrity of the affected component(s).

CR-3 has performed inspections of the RCPB and the RVH during previous refueling outages using volumetric, surface, and visual examination techniques. The visual examinations include direct observation and indirect observation, for leakage and Boric Acid residue. The direct inspection of the RVH is conducted through the access openings in the Control Rod Drive Service Structure (CRDSS) and is a bare metal inspection. Direct examinations are also performed on other Alloy 600 components and bolted connections.

Indirect inspection is performed through the observation of evidence of leakage; i.e., signs of boric acid accumulation. These visual inspections meet the requirements of Section XI Table IWB-2500-1 and IWB-3522. The visual inspections also meet the requirements of NRC Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants." Compliance with the requirements of Section XI is implemented through the CR-3 Inservice Inspection Program. If the VT-2 examinations detect the conditions described in IWB-3522.1, as not meeting the acceptance of IWB-3142, then the corrective actions required would be performed in accordance with IWA-5250 (Corrective Measures) and the CR-3 CAP. During Refueling Outage 12 (2001), one CRDM nozzle was identified and confirmed as leaking from the visual inspections of the Reactor Vessel Head (RVH). The CRDM nozzle was repaired prior to restart from the refueling outage. No degradation of the RVH carbon steel was identified.

CR-3 ITS 3.4.12, "RCS Operational LEAKAGE," LCO 3.4.12a states, "RCS operational LEAKAGE shall be limited to: No pressure boundary LEAKAGE."

U. S. Nuclear Regulatory Commission Page 15 of 15 3F0103-03 Attachment Monitoring and various leakage detection systems are available that provide diverse methods of detection of unidentified leakage to the plant operator to ensure appropriate corrective actions are taken in accordance with ITS.

When the unidentified plant leakage approaches the plant administrative limits, appropriate actions will be taken to identify leakage sources to ensure that further degradation of the RCPB does not continue. Discovery of RCPB leakage would require the plant to shutdown.

Visual inspections conducted during refueling outages provide the opportunity to access areas/components within the plant that are normally not accessible during plant operations.

The program has assured that no significant wastage has occurred as a result of boric acid corrosion. The program will continue being evaluated and enhanced, as needed, incorporating industry experience and best practices.

A round-robin self-assessment of all three Progress Energy PWR's boric acid corrosion control programs has been performed. The results of the self-assessment will be used to enhance some aspects of the programs to ensure high standards, requirements, and operating experience are consistently implemented at each site.

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