ML18153B110
| ML18153B110 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/20/1994 |
| From: | Buckley B Office of Nuclear Reactor Regulation |
| To: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| TAC-M89085, NUDOCS 9410260140 | |
| Download: ML18153B110 (6) | |
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October-20;; 1994 Mr. J.P. O'Hanlon Senior Vice President*- Nuclear Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, VA 23060
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON THE SURRY POWER STATION, UNIT 2, THIRD IO-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR RELIEF (TAC NO. M89085)
Dear Mr. O'Hanlon:
The NRC staff, with assistance from our contractor, Idaho National Engineering Laboratory (INEL), is evaluating your March 18, 1994, submittal on the Surry, Unit 2 Third IO-year Inservice Inspection Program.
Based on our evaluation, we find that additional information is required in order to continue our review.
It is requested that you provide responses to the questions delineated in the enclosure within 30 days of receipt of this letter.
In addition, to expedite the review process, please provide a copy of your response to our contractor, INEL, at the following address:
Boyd W. Brown INEL Research Center 2151 North Boulevard P.O. Box 1625 Idaho Falls, Idaho 83415-2209 This requirement affects fewer than 10 respondents and, therefore, is not subject to Office of Management and Budget Review under P.L.96-511.
Sincerely, (Original Signed By)
Bart C. Buckley, Senior Project Manager Project Directorate II-2 Docket No. 50-281
Enclosure:
As stated cc w/enclosure: See next page Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Distribution r::oocl<et. Fn e PUBLIC PDll-2 RF SVarga JSwol i nski OGC ACRS (10)
DVerelli, RII Document Name: C:\\AUTOS\\WPDOCS\\SURRY\\RAI2.IST To receive a copy of this document, indicate in attachment/enclosure "E" = Copy with attachme e box:
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Mr. J. P. O'Hanlon Virginia Electric and Power Company cc:
Michael W. Maupin, Esq.
Hunton and Williams Riverfront Plaza, East Tower 951 E. Byrd Street Richmond, Virginia 23219 Mr. Michael R. Kansler, Manager Surry Power Station Post Office Box 315 Surry, Virginia 23883 Senior Resident Inspector Surry Power Station U.S. Nuclear Regulatory Commission 5850 Hog Island Road Surry, Virginia 23883 Chairman Board of Supervisors of Surry County Surry County Courthouse Surry, Virginia 23683 Dr. W. T. Lough Virginia State Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23209 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 Robert B. Strobe, M.D., M.P.H.
State Health Commissioner Office of the Commissioner Virginia Department of Health P.O. Box 2448 Richmond, Virginia 23218 Surry Power Station Attorney General Supreme Court Building 101 North 8th Street Richmond, Virginia 23219 Mr. M. L. Bowling, Manager Nuclear Licensing & Programs Innsbrook Technical Center Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, Virginia 23060
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VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2 DOCKET NUMBER 50-281 MATERIALS AND CHEMICAL ENGINEERING BRANCH DIVISION OF ENGINEERING Reguest for Additional Information - Third IO-Year Interval Inservice Inspection Program Plan
- 1. Scope/Status of Review Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the requirements, except design and access provisions and preservice examination requirements, set forth in the ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.SSa(b) on the date 12 months prior to the start of a successive 120-month interval, subject to the limitations and modifications listed therein.
The components (including supports) may meet requirements set forth in subsequent editions and addenda of the Code that are incorporated by reference in 10 CFR 50.SSa(b) subject to the limitations and modifications listed therein and subject to Nuclear Regulatory Commission (NRC) approval.
The licensee, Virginia Electric and Power Company, has prepared the Surry Power Station, Unit 2, Third Ten-Year Interval lnservice Inspection Program, Revision 0, to meet the requirements of the 1989 Edition of the ASME Code,Section XI.
As required by 10 CFR 50.SS{g)(S), if the licensee determines that certain Code examination requirements are impractical and requests relief, the licensee shall submit information to the NRC to support that determination.
The staff has reviewed the available information in the Surry Power Station, Unit 2, Third Ten-Year Interval Inservice Inspection Program, Revision 0, submitted to the NRC by letter dated March 18, 1994.
- 2.
Additional Information Reguired Based on the review of the licensee's submittal, the staff has concluded that the following information and/or clarification is required to complete the review of the inservice inspection program plan.
A.
Address the degree of compliance with augmented examinations that have been established by the NRC when added assurance of structural reliability is deemed necessary.
Examples of documents that address
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augmented examinations that may be applicable based on licensee commitments are listed below:
(1)
Branch Technical Position MEB 3-1, High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment; (2)
Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations; B.
The Code of Federal Regulations, Part 10, 50.55a(g)(6)(ii)(A),
requires that all licensees must augment their reactor vessel examinations by implementing once, during the inservice inspection interval in effect on September 8, 1992, the examination requirements for reactor vessel shell welds specified in Item 81.10 of Examination Category 8-A of the 1989 Code.
In addition, all previously granted relief for Item 81.10, Examination Category 8-A, for the interval in effect on September 8, 1992, is revoked by the new regulation.
For licensees with fewer than 40 months remaining in the interval on the effective date, deferral of the augmented examination is permissible with the conditions stated in the regulations.
Based on the effective date of the subject regulation and the May 10, 1994, starting date of the third 10-year interval of the Surry Power Station, Unit 2, please provide the staff with the projected schedule for this augmented examination and a technical discussion describing how it will be implemented at Surry Power Station, Unit 2, during the third interval. Describe the intended approach and any specialized techniques or equipment that will be used to complete the required augmented examinations.
Include an estimate on the percentage of volumetric coverage that can and will be achieved.
C.
Portions of Code Class 2 piping welds in the Residual Heat Removal, Emergency Core Cooling, and Containment Heat Removal systems are critical to the safe shutdown of the plant. It has been recognized that current Code examination requirements exclude selection of thin-wall_ piping welds (<3/8 inch) in the subject systems.
As a result, flaws in thin-wall piping welds would not be detected until through-wall leakage occurs.
In the review of the licensee's program, it has been noted that Class 2 piping welds <3/8 inch are included in the total Class 2 piping weld population but are excluded from examinations.
Considering the safety significance of the subject systems, describe your plans for performing volumetric examinations on a sample of thin-wall piping welds to assure the continued integrity of the subject systems.
D.
In Request for Relief SR-002, the licensee requested relief from performing a volumetric examination of the steam generator, primary side nozzle inner radius sections. The licensee has deemed the volumetric examination impractical due to factors that include component geometry, long metal paths, and material attenuation.
However, in the conclusion the licensee states that to perform an
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3 examination, a mockup would be required and training provided to examination personnel.
Many ISi examinations require mockups as well as additional training for examination personnel to meet Code requirements and to ensure a meaningful examination.
The licensee's position that the subject examinations are impractical is not supported by the basis for relief provided.
Provide further discussion to support a conclusion that the Code-required examinations are impractical.
E.
In Request for Relief SR-003, the licensee described the limiting factors for the examination of the pressurizer surge nozzle inside radius section and concluded that any examination of this nozzle would be a "best effort". A "best effort" volumetric examination would provide a level of assurance that flaws are not initiating in the inside radius section. Discuss the extent of Code coverage that can and will be achieved by performing a "best effort" examination.
F.
For Request for Relief SR-005, the licensee proposed the use of existing ultrasonic calibration blocks without modifying them to satisfy current Code requirements. It is expected that licensees will meet the requirements for the applicable interval by upgrading calibration block designs.
Provide a list of calibration blocks and describe where the existing designs differ from the applicable Code requirements for the third interval. Describe how the existing calibration blocks will provide the same level of examination quality as those designs currently required by the Code in effect.
G.
Request for Relief 1 addresses the pressure test of the piping between Valves MOV-2700 and MOV-2701.
The licensee stated that valves MOV-2700 and MOV-2701 are closed for the pressure test of the Class 1 side to avoid over pressurization of the Class 2 side.
Explain why MOV-2701 is not an acceptable pressure boundary for the pressurization of the segment of line between the subject valves, which is required to be tested at the higher of the operating pressures when implementing Code Case N-498.
Describe how pressurization of the Class 1 segment of line between the subject valves at the lower test pressure provides the same level of assurance of structural integrity as the Code-required test pressure.
H.
In Request for Relief 4, is the licensee proposing to use the lower pressure associated with the auxiliary feedwater pump on the high pressure side and take credit for the pressure test on both sides?
If so, how does this lower pressure provide reasonable assurance of component integrity for the high pressure side?
I.
Recent incidences of degraded bolting have reinforced the requirement to remove a bolt for a VT-3 visual examination as part of the evaluation process. (
Reference:
NRC Event 26899, dated 03/08/94, and 26992, dated 03/25/94).
Because degradation rates cannot be reliably predicted and bolting material records may not be
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accurate, direct visual examination and invnediate corrective action for leakage at bolted connections is warranted.
For Request for Relief 5, it is not apparent that the licensee intends to remove at least one bolt nearest the source of leakage for a VT-3 visual examination as part of each evaluation.
Verify that at least one bolt, closest to the source of leakage, will be removed for a VT-3 visual examination for each leakage occurrence as part of the evaluation.
J.
For Code Class I integral attachment welds to p1p1ng, pumps, and valves, the Code does not require examinations for the third and fourth interval when implementing Inspection Program B.
Examination of integral attachments in Code Class 2 and 3 systems is required in the third and fourth interval.
ASME Code Case N-509 (approved November 25, 1992 by ASME) provides for continued examination of Class I integral attachments for the life of the plant as well as readjustments in the sample inspection requirements for Code Class 2 and 3.
Describe your plans with respect to examination of Code Class I integral attachment welds during the third inspection interval or implementation this Code Case.
K.
Verify that no additional requests for relief are required at this time.
If additional relief requests are required, the licensee should submit them for staff review.
The schedule for timely completion of this review requires that the licensee provide, by the requested date, the above requested information and/or clarification with regard to the Surry Power Station Unit 2, Third Ten-Year Interval Inservice Inspection Program, Revision 0.