ML023260180

RAI, Bulletin 2002-01, Reactor Pressure Vessel Head Degradation & Reactor Coolant Pressure Boundary Integrity, 15-Day & 60-Day Responses

ML023260180
Person / Time
Site:	Palo Verde
Issue date:	11/21/2002
From:	Donohew J NRC/NRR/DLPM/LPD4
To:	Overbeck G
Donohew J N, NRR/DLPM,415-1307
References
BL-02-001, TAC MB4563, TAC MB4564, TAC MB4565
Download: ML023260180 (6)

Text

November 21, 2002 Mr. Gregg R. Overbeck Senior Vice President, Nuclear Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034

SUBJECT:

BULLETIN 2002-01, "REACTOR PRESSURE VESSEL HEAD DEGRADATION AND REACTOR COOLANT PRESSURE BOUNDARY INTEGRITY," 15-DAY AND 60-DAY RESPONSES FOR PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 - REQUEST FOR ADDITIONAL INFORMATION (TAC NOS. MB4563, MB4564, AND MB4565)

Dear Mr. Overbeck:

On March 18, 2002, the Nuclear Regulatory Commission (NRC) issued Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity," to all holders of operating licenses for pressurized water reactors (PWRs) requesting that addressees provide information related to the structural integrity of the reactor coolant pressure boundary, including the reactor pressure vessel head, and the extent to which inspections have been undertaken to satisfy applicable regulatory requirements. Within 15 days of the date of this bulletin, the addressees were requested to respond to Items 1.A.

through E. of the bulletin, or to provide a written response to the NRC in accordance with the provisions of Title 10, Section 50.54(f), of the Code of Federal Regulations (10 CFR 50.54(f)) if they are unable to provide the information or they cannot meet the requested completion dates.

Within 60 days of the date of the bulletin, the addresses were required to submit to the NRC the following information related to the reactor coolant pressure boundary (RCPB) other than the reactor pressure vessel (RPV) head:

The basis for concluding that your boric acid inspection program is providing reasonable assurance of compliance with the applicable regulatory requirements discussed in Generic Letter 88-05 and this bulletin. If a documented basis does not exist, provide your plans, if any, for a review of your programs.

Based on the review of your 15-day bulletin response dated April 3, 2002 (102-04681), the NRC staff finds that you have provided the requested information, and has concluded that your plant does not appear to have conditions similar to those which lead to the degradation at Davis-Besse.

The NRC staff has evaluated licensees 60-day responses to Bulletin 2002-01 concerning the rest of the RCPB. The response for Palo Verde, Units 1, 2, and 3 is dated May 17, 2002 (102-04702). The staff concluded that most of the licensees 60-day responses lacked specificity. Therefore, the staff could not complete its review of the boric acid corrosion control (BACC) programs in light of the lessons learned from the Davis-Besse event. The information request in Bulletin 2002-01 may not have been sufficiently focused, which, in part, may explain the lack of clarity in the licensees 60-day responses. The staffs review of all licensees 60-day responses provided the basis for development of the questions in this request for additional

G. Overbeck information (RAI). Licensees are expected to provide responses in sufficient details to facilitate a comprehensive staff review of their BACC programs.

The NRC is not imposing new requirements through the issuance of Bulletin 2002-01, or this RAI. The NRC staffs review of the information collected will be used as part of the decisionmaking process regarding possible changes to the NRCs regulation and inspection of BACC programs. The NRC staff has, however, concluded that a comprehensive BACC program would exceed the current American Society of Mechanical Engineers Code requirements; and would include, but is not limited to, the following:

1. The BACC program must address, in detail, the scope, extent of coverage, degree of insulation removal, and frequency of examination for materials susceptible to boric acid corrosion. The BACC program would also ensure that any boric acid leakage is identified before significant degradation occurs which may challenge structural integrity.

a. The scope should include all components susceptible to boric acid corrosion (BAC) and identify the type of inspection(s) performed (e.g., VT-2 or VT-3 examination).

b. The technical basis for any deviations from inspection of susceptible materials and mechanical joints must be clearly documented.

c. As stated in Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," the BACC program should identify the principal locations where leaks that are smaller than the allowable technical specification limit have the potential to cause degradation of the primary pressure boundary by boric acid corrosion. Particular consideration should be given to identifying those locations where conditions exist that could cause high concentrations of boric acid on pressure boundary surface, or locations that are susceptible to primary water stress corrosion cracking (Alloy 600 base metal and dissimilar metal Alloy 82/182 welds), or susceptible to leakage (e.g., valve packing, flange gaskets).

d. For inaccessible components (e.g., buried components, components within rooms, vaults, etc.) the degree of inaccessibility, and the type of inspection that would be effective for examination of the area must be clearly defined. In addition, identify any leakage detection systems that are being used to detect potential leakage from components in inaccessible areas.

e. The technical basis for the frequency of implementing the BACC program must be clearly documented.

2. The examiners would be VT-2 qualified at a minimum, and would be trained to recognize that very small volumes of boric acid leakage could be indicative of significant corrosion.

G. Overbeck 3. The BACC program would ensure that any boric acid leakage is identified before significant degradation occurs which may challenge structural integrity. If observed leakage from mechanical joints is not determined to be acceptable, the appropriate corrective actions must be taken to ensure structural integrity. Evaluation criteria and procedures for structural integrity assessments must be specified. The applicable acceptance standards and its bases must also be identified.

4. Leakage from mechanical joints (e.g., bolted connections) that is determined to be acceptable for continued operation must be inspected and monitored in order to trend/evaluate changes in leakage. The bases for acceptability must be documented.

Any evaluation for continued service should include consideration of corrosion mechanisms and corrosion rates. If boric acid residues are detected on components, the leakage source shall be located by removal of insulation, as necessary.

Identification of the type of insulation and any limitations concerning its removal should be addressed in the BACC program.

5. Leakage identified outside of inspections for BAC should be integrated into the BACC program.

6. Licensees would routinely review and update the BACC program in light of plant specific and industry experience, monitoring and trending of past leakage, and proper documentation of boric acid evaluations to aid in determination of recurring conditions and root cause of leakage. New industry information should be integrated in a consistent manner such that revised procedures are clear and concise.

Please consider the above attributes in providing your responses to the enclosed RAI. This request was discussed with your staff on November 20, 2002, and it was agreed that a response to the enclosed RAI would be provided on or before January 31, 2003. We will complete our review of your responses to the bulletin after we have reviewed the responses to Items 2 and 3. If you have any questions, please contact me at 301-415-1307.

Sincerely,

/RA/

Jack Donohew, Senior Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. STN 50-528, STN 50-529, and STN 50-530

Enclosure:

Request for Additional Information cc w/encl: See next page

G. Overbeck 3. The BACC program would ensure that any boric acid leakage is identified before significant degradation occurs which may challenge structural integrity. If observed leakage from mechanical joints is not determined to be acceptable, the appropriate corrective actions must be taken to ensure structural integrity. Evaluation criteria and procedures for structural integrity assessments must be specified. The applicable acceptance standards and its bases must also be identified.

4. Leakage from mechanical joints (e.g., bolted connections) that is determined to be acceptable for continued operation must be inspected and monitored in order to trend/evaluate changes in leakage. The bases for acceptability must be documented.

Any evaluation for continued service should include consideration of corrosion mechanisms and corrosion rates. If boric acid residues are detected on components, the leakage source shall be located by removal of insulation, as necessary.

Identification of the type of insulation and any limitations concerning its removal should be addressed in the BACC program.

5. Leakage identified outside of inspections for BAC should be integrated into the BACC program.

6. Licensees would routinely review and update the BACC program in light of plant specific and industry experience, monitoring and trending of past leakage, and proper documentation of boric acid evaluations to aid in determination of recurring conditions and root cause of leakage. New industry information should be integrated in a consistent manner such that revised procedures are clear and concise.

Please consider the above attributes in providing your responses to the enclosed RAI. This request was discussed with your staff on November 20, 2002, and it was agreed that a response to the enclosed RAI would be provided on or before January 31, 2003. We will complete our review of your responses to the bulletin after we have reviewed the responses to Items 2 and 3. If you have any questions, please contact me at 301-415-1307.

Sincerely,

/RA/

Jack Donohew, Senior Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation DISTRIBUTION:

Docket Nos. STN 50-528, STN 50-529, PUBLIC and STN 50-530 PDIV-2 Reading EMCB Reading RidsNrrDlpmLpdiv (WRuland)

Enclosure:

Request for Additional Information SDembek RidsNrrPMJDonohew WBateman cc w/encl: See next page RidsNrrLAMMcAllister SBloom RidsOgcRp SCoffin RidsAcrsAcnwMailCenter ESullivan ACCESSION NO. ML02 NRR-088 RidsRgn4MailCenter (LSmith)

OFFICE PDIV-2/PM PDIV-2/LA EMCB PDIV-2/SC NAME JDonohew:rkb EPeyton SBloom SDembek DATE 11/20/02 11/20/02 11/21/02 11/21/02 DOCUMENT NAME: C:\ORPCheckout\FileNET\ML023260180.wpd OFFICIAL RECORD COPY

REQUEST FOR ADDITIONAL INFORMATION REGARDING BORIC ACID CORROSION CONTROL PROGRAMS PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 DOCKET NOS. STN 50-528, STN 50-529, AND STN 50-530 The format provided in Table A may be used to respond to the following RAIs:

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 (RCPB). Include specific discussion of inspection of locations where reactor coolant leaks have the potential to come in contact with and degrade the subject material (e.g., reactor pressure vessel (RPV) bottom head).

2. Provide the technical basis for 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 corrosion cracking (Alloy 600 base metal and dissimilar metal Alloy 82/182 welds). Identify the type of insulation for each component examined, as well as any limitations to 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 acid deposits) are found.

3. Describe the technical basis for the extent and frequency of walkdowns and the method for evaluating the potential for leakage in inaccessible areas. In addition, describe the degree of inaccessibility, and identify any leakage detection systems that are being used to detect potential leakage from components in inaccessible areas.

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 was established to make such a determination. Provide the technical basis used to establish the acceptance criteria. In addition,

a. if observed leakage is determined to be acceptable for continued operation, describe what inspection/monitoring actions are taken to trend/evaluate changes in leakage, or

b. if observed leakage is not determined to be acceptable, describe what corrective actions are taken to address the leakage.

5. Explain the capabilities of your program to detect the low levels of reactor coolant 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 has the potential for causing boric acid corrosion. 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.

6. Explain the capabilities of your program to detect the low levels of reactor coolant pressure boundary leakage that may result from through-wall cracking in certain components and configurations for other small diameter nozzles. Low levels of leakage may call into question reliance on visual detection techniques or installed leakage detection instrumentation, but has 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.

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.

8. Provide a summary of recommendations made by your reactor vendor on visual inspections of nozzles with Alloy 600/82/182 material, actions you have taken or plan to take regarding vendor recommendations, and the basis for any recommendations that are not followed.

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 Federal Regulations (10 CFR), 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, paragraph IWA-5250(b) on corrective actions. Include a description of the procedures used to implement the corrective actions.

Table A. Template for Response to RAIs Component Inspection Personnel Extent of Frequency Degree of Insulation Corrective Techniques Qualifications Coverage Removal/Insulation Action Type

Palo Verde Generating Station, Units 1, 2, and 3 cc:

Mr. Steve Olea Mr. David Summers Arizona Corporation Commission Public Service Company of New Mexico 1200 W. Washington Street 414 Silver SW, #1206 Phoenix, AZ 85007 Albuquerque, NM 87102 Douglas Kent Porter Mr. Jarlath Curran Senior Counsel Southern California Edison Company Southern California Edison Company 5000 Pacific Coast Highway Building DIN Law Department, Generation Resources San Clemente, CA 92672 P.O. Box 800 Rosemead, CA 91770 Mr. Robert Henry Salt River Project Senior Resident Inspector 6504 East Thomas Road U.S. Nuclear Regulatory Commission Scottsdale, AZ 85251 P.O. Box 40 Buckeye, AZ 85326 Terry Bassham, Esq.

General Counsel Regional Administrator, Region IV El Paso Electric Company U.S. Nuclear Regulatory Commission 123 W. Mills Harris Tower & Pavillion El Paso, TX 79901 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 Mr. John Schumann Los Angeles Department of Water & Power Chairman Southern California Public Power Authority Maricopa County Board of Supervisors P.O. Box 51111, Room 1255-C 301 W. Jefferson, 10th Floor Los Angeles, CA 90051-0100 Phoenix, AZ 85003 Brian Almon Mr. Aubrey V. Godwin, Director Public Utility Commission Arizona Radiation Regulatory Agency William B. Travis Building 4814 South 40 Street P.O. Box 13326 Phoenix, AZ 85040 1701 North Congress Avenue Austin, TX 78701-3326 Mr. Craig K. Seaman, Director Regulatory Affairs/Nuclear Assurance Palo Verde Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072-2034 Mr. Hector R. Puente Vice President, Power Generation El Paso Electric Company 2702 N. Third Street, Suite 3040 Phoenix, AZ 85004