ML18151A901

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ISI Program Interval 3,940510- 040509, Rev 0
ML18151A901
Person / Time
Site: Surry Dominion icon.png
Issue date: 03/03/1994
From: Mcneill A
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18151A182 List:
References
PROC-940303, NUDOCS 9403230233
Download: ML18151A901 (71)
v • d • e


Text

,:*. . . .:

  • VIRGINIA POWER INSERVICE INSPECTION PROGRAM CHANGE / ISSUE APPROVAL FORM STATION: SURRY UNIT NO.: 2 INTERVAL NO. : 3 SUBMITTAL TITLE: INSERVICE INSPECTION PROGRAM CHANGE NO.: SECTION NO. : REVISION: 0 (IF APPL! CABLE)

DESCRIPTION OF CHANGES:

INITIAL ISSUE OF THE INTERVAL 3 INSERVICE INSPECTION PROGRAM FOR COMPONENTS AND COMPONENT SUPPORTS (ATTACH PAGE(S) SHOWING EXACT WORDING OF PROPOSED CHANGE)

REASON FOR CHANGE:

      • ORIGINATED BY: ALEX MCNEILL DEPARTMENT: ISI/NDE AND ENGINEERING PROGRAMS SUPERVISOR - ISI/NDE INSPECTION PROGRAMS DATE: 2 /2 2/94 REVIEWED AND APPROVED: DATE: oz/zz.,/1c./

SUPERVISOR - ISI/NDE INSPECTION PROGRAMS (STATION)

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DATE: 3/3/9,/

ANII ( S T A T I O ~

REVIEWED: {?. DATE:

STATION NUCLEAR SAFE'rf AND OPERATING COMMITTEE (SNSOC)

REVIEWED AND vJ-z~,i/ :A APPROVED:~~~~-~-m~-/\.A/~"--"-'~~~~~~~~~~ DATE:

AFTER APPROVAL, Rzjt'URN TO: Joyce A. Lindsay, IN-3NW ====-

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  • Virginia Power, 5000 Dominion Blvd., Glen Allen, VA DISTRIBUTION NO.: DISTRIBUTION DATE:

23060 FORM?-15.\.1£'_5__~ --_,

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VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2 INSERVICE INSPECTION PROGRAM INTERVAL 3

  • MAY 10, 1994 - MAY 9, 2004 REVISION 0
    • S2-001I3 i

Rev. 0 12/93

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2 INSERVICE INSPECTION PROGRAM INTERVAL 3 MAY 10, 1994 TO MAY 9, 2004 REVISION 0 DISTRIBUTION RECORD DISTRIBUTION DATE DISTRIBUTION DATE NO. ENTERED INITIAL NO. ENTERED INITIAL 1 21 2 22 3 23 4 24 5 25 6 26 7 27 8 28

  • 9 10 11 29 30 31 12 32 13 33 14 34 15 35 16 36 17 37 18 38 19 39 20 40 Future distributions to this report will be accompanied by two copies of the distribution memo, one of which is to be returned per instructions thereon, and the other one is to be filed sequentially behind this page giving you a record of the contents of each distribution.

S2-001I3 Rev. 0 ii 12/93

  • ABSTRACT VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2 INTERVAL 3 MAY 10, 1994 TO MAY 9, 2004 In accordance with 10CFR50.55a the Surry Unit 2 Inservice Inspection and Testing Program was updated to meet the requirements of ASME Section XI, 1989 Edition. Stearn generator inspections will continue to be inspected under Plant Technical Specifications. Specific reliefs are requested in accordance with 10CFR50.55a(g) (5) (iii).

The interval for which this program is applicable commenced on May 10, 1994, and will end on May 9, 2004.

The Inservice Inspection Program was developed employing 10CFR50 and Reg. Guide 1.26. Quality Groups A, B, and Care the same as ASME Classes 1, 2, and 3 respectively .

  • The list of drawings on the following pages identifies the drawings used in developing the program.

Section 1 introduces the Inservice Inspection Program.

Section 2 describes the Class 1, 2, and 3 component Inservice Inspection Program developed in accordance with Subsections IWB, IWC, and IWD of ASME Section XI.

Section 3 describes the Inservice Inspection Program for component supports .

  • S2-001I3 iii Rev. 0 12/93
  • SECTION TITLE PAGE TABLE OF CONTENTS PAGE NO. REVISION i Assignment i 0 ii Distribution Record ii 0 iii Abstract iii 0 iv Table of Contents iv 0 V List of Drawings V to vii 0 1 Introduction 1-1 to 1-2 0 2 Inservice Inspection Program 2-1 to 2-36 0 for Components 3 Inservice Inspection Program 3-1 to 3-3 0 for Component Supports
  • 4 Interim Relief Requests and 4-1 to 4-2 0 Miscellaneous Documentation & Documents
  • S2-001I3 iv Rev. 0 12/93
  • UNIT 2 11548-CB-L&S-3 LIST OF DRAWINGS CB DRAWINGS Legend and Symbols UNIT 2 CBB DRAWINGS 11548-CBB-006A-3 Air Cooling and Purging System 11548-CBB-047B-3 Fire Protection System UNIT 2 CBM DRAWINGS 11548-CBM-064A-3 Main Steam System 11548-CBM-064B-3 Steam Generator Nitrogen Connection System 11548-CBM-066A-3 Auxiliary Steam and Air Removal System
  • 11548-CBM-067A-3 11548-CBM-068A-3 11548-CBM-071A-3 Condensate System Feedwater System Circulating and Service Water System 11548-CBM-071B-3 Circulating and Service Water System 11548-CBM-072A-3 Component Cooling System 11548-CBM-072B-3 Component Cooling System 11548-CBM-072C-3 Component Cooling System 11548-CBM-075B-3 Compressed Air System 11548-CBM-075C-3 Compressed Air System 11548-CBM-075E-3 Compressed Air System
  • S2-001I3 V

Rev. 0 12/93

  • UNIT 2 11548-CBM-075J-3 LIST OF DRAWINGS CONT.

CBM DRAWINGS Containment Instrument Air System 11548-CBM-082A-3 Sampling System 11548-CBM-083A-3 Vent and Drains System 11548-CBM-083B-3 Vent and-Drains System 11548-CBM-084A-3 Containment Spray System 11548-CBM-084B-3 Recirculation Spray System 11548-CBM-085A-3 Containment Vacuum and Leakage Monitoring System 11548-CBM-086A-3 Reactor Coolant System 11548-CBM-086B-3 Reactor Coolant System 11548-CBM-087A-3 Residual Heat Removal System

  • 11548-CBM-088A-3 11548-CBM-088B-3 11548-CBM-088C-3 Chemical and Volume Control System Chemical and Volume Control System Chemical and Volume Control System 11548-CBM-089A-3 Safety Injection System 11548-CBM-089B-3 Safety Injection System 11548-CBM-llSA-3 Reactor Cavity Purification System 11548-CBM-124A-3 Steam Generator Blowdown and Recirc. and Transfer System 11548-CBM-lJOB-3 Containment Particulate System Additionally, several Unit 1 drawings contain Unit 2 Components.

These follow:

11448-CBM-068A-3 Feedwater System 11448-CBM-071B-3 Circulating and Service Water System

  • S2-001I3 vi Rev. 0 12/93
  • 11448-CBM-0710-3 11448-CBM-072C-3 11448-CBM-0720-3 Circulating and Service Water System Component Cooling System Component Cooling System

(

11448-CBM-072E-3 Component Cooling System 11448-CBM-072G-3 Component Cooling System 11448-CBM-082A-3 Sampling System 11448-CBM-OSSA-3 Chemical and Volume Control System 11448-CBM-090C-3 Containment Hydrogen Analyzer System

  • S2-001I3 vii Rev. 0 12/93

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT2

  • INSERVICE INSPECTION PROGRAM
  • S2-001I3 1-1 Rev. 0 12/93

1.0 INTRODUCTION

1.1 GENERAL INFORMATION Surry Power Station Unit 2 is a Pressurized Water Reactor located on Gravel Neck and adjacent to the James River in Surry County, Virginia. The plant employs a Westinghouse Electric Corp. Nuclear Steam System.

The Inservice Inspection (ISI) Program for Surry Nuclear Power Station Unit 2 was developed in compliance with the rules and regulations of 10CFR50.55a and Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition. Where these rules are determined to be impractical, specific relief is requested in writing.

The Inservice Inspection Program for Class 1, 2, and 3 Components and Components Supports is applicable for the ten year interval beginning May 10, 1994 and ending May 9, 2004. This interval is the third inspection interval for Surry Unit 2.

1.2 SYSTEM CLASSIFICATION

  • The construction permits for Surry Units 1 and 2 were issued on June 25, 1968. At that time the ASME Boiler and Pressure Vessel Code covered only pressure vessels.

Piping, pumps, and valves were built primarily to the rules of USAS B31.l. Essentially, Surry Power Station was designed and constructed prior to the origination of the ASME Code classifications named Class 1, 2, and

3. Therefore the system classifications used as a basis for the Inservice Inspection and Testing Programs are based on the requirements set forth in 10CFR50 and Regulatory Guide 1.26. Pursuant to 10CFR50.55a paragraph (g) (1), inservice inspection requirements of Section XI of the ASME Code are then assigned to these components, within the constraints of existing plant design.

Classification boundary drawings (CBM's) documenting the system classifications were developed to aid in the review and implementation of the subject programs .

  • S2-001I3 1-2 Rev. 0 12/93

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT2

  • INSERVICE INSPECTION PROGRAM FOR COMPONENTS
  • S2-002I3 2-1 Rev. 0 12/93
  • SECTION 2 TABLE OF CONTENTS 2.0 INSERVICE INSPECTION PROGRAM FOR COMPONENTS 2.1 PROGRAM DESCRIPTION 2.2 PROGRAM

SUMMARY

2.3 RELIEF REQUESTS

  • S2-002I3 2-2 Rev. 0 12/93
  • 2.0 INSERVICE INSPECTION PROGRAM FOR COMPONENTS 2.1 PROGRAM DESCRIPTION 2.1.1 The Inservice Inspection Program for Class 1, 2 and 3 components meets the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition. Where these requirements are determined to be impractical, specific requests for relief have been written and included in this section.

2.1.2 The 1989 Edition of ASME Section XI includes subsections IWE and IWL for Class MC and Class cc components. This program currently excludes these components as not being within the scope of 10CFR50.55a at this time.

2.1.3 Repairs and replacements shall be in accordance with Station Administrative Procedures, which assure Code Requirements are satisfied.

2.2 PROGRAM

SUMMARY

2.2.1 The Inservice Inspection Program for Surry Unit 2 utilizes the tables IWB-2500-1, IWC-2500-1, and IWD-2500-1 of the 1989 Edition of the ASME Section XI Code. Components selected for examination are identified in the Inservice Inspection Program Plan.

2. 2. 2 Categories B-E, B-P, C-H, D-A, D-B, and D-C constitute the scheduled 10-year interval pressure testing requirements. Additionally, Code Case N-498, see below, will be followed.

The following guidance will be utilized programmatically concerning pressure testing.

a) Pressure tests which meet the pressure, temperature, hold time, and VT-2 examination requirements of IWA-5000, IWB-5000, IWC-5000, IWD-5000, or Code Case N-498 as applicable will be treated as complete for inspection scheduling purposes.

b) Visual VT-2 examinations of tubes within the steam generators or other heat exchangers will not be conducted, as it is not considered within the scope of ASME Section XI.

  • S2-002I3 2-3 Rev. 0 12/93
  • c) Pressure tests required for ASME Section XI repairs or replacements will be conducted in accordance with the station's repair replacement program as documented at the station.

2.2.3 The following Code Cases are in use at Surry Power Station Unit 2, subject to modification per Reg. Guide 1.147.

  • code case N-401 Eddy Current Examination,Section XI Division 1.

Code case N-402 - Eddy Current Calibration Standard Material,Section XI, DivTsion 1.

Code Case N-416 - Alternative Rules for Hydrostatic Testing of Repair or Replacement of Class 2 Piping,Section XI, Division 1.

Code Case N-435 Alternative Examination Requirements for Vessels with Wall thickness 2 in. or less,Section XI, Division 1.

Code Case N-437 ~ Use of Digital Readout and Digital Measurement Devices for Performing Pressure TestsSection XI, Division 1.

Code Case N-460 - Alternative Examination Coverage for Class 1 and Class 2 Welds,Section XI, Division 1.

Code Case N-461 - Alternative Rules for Piping Calibration Block Thickness,Section XI, Division 1.

Code Case N-481 - Alternate Examination Requirements for Cast Austenitic Pump Casings,Section XI, Division 1.

Code Case N-489 - Alternative Rules for Level III NDE Qualification Examinations,Section XI, Division 1 .

S2-002I3 Rev. 0 2-4 12/93

2. 2. 4 Isometric Drawings (WMKS) do not specifically identify underground piping welds, however weld selection in the ISI Plan has included the number of underground welds in the percentages selected.

2.2.5 New High Pressure Safety Injection Isometric Drawings (WMKS) were required for the third interval update. These new drawings were based upon drawings which do not give precise support information. These drawings will require further development to incorporate the required component support information. The drawing information and required component support examinations will be completed prior to the end of the third interval.

2.2.6 The requirements of Regulatory Guide 1.150, Rev. 1, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, will be followed.

2.2.7 Augmented inspections coordinated by the Inservice Inspection Group are identified within the Inservice Inspection Program Plan.

2.3 RELIEF REQUESTS 2.3.1 Nondestructive examination (NDE) relief requests are presented in a numeric sequence with a SR prefix. Pressure tests relief requests are presented in a numeric sequence with no prefix .

  • S2-002I3 2-5 Rev. 0 12/93
  • I.

Weld #

RELIEF REQUEST SR-001 IDENTIFICATION OF COMPONENTS Drawing # Class 1-0lDM 11548-WMKS-RC-10-l 1 l-17DM 11548-WMKS-RC-10-l 1 1-0lDM 11548-WMKS-RC-ll-l 1 l-17DM 11548-WMKS-RC-ll-l 1 1-0lDM 11548-WMKS-RC-12-l 1 l-17DM 11548-WMKS-RC-12-l 1 The reactor vessel nozzle-to-safe end butt welds II. IMPRACTICAL CODE REQUIREMENTS Category B-F, Item B5.10 requires that a volumetric and surface examination be performed. The surface examination must be performed from the outside diameter of the reactor vessel nozzle-to-safe end welds.

III. BASIS FOR RELIEF The outside diameter volumetric examination would be extremely difficult to perform. Access to the area is restricted by permanent neutron shielding and support structures. Any planned removal to provide access is made difficult by the relatively small sandplug access on the floor of the refueling cavity. The difficult access restrictions are also complicated by the anticipated high dose level. The general area estimate around the perimeter of the vessel and nozzle area is 1000 to 2000 MR/HR. The contact estimate ranges from 4000 to 8000 MR/HR in the vicinity of the sliding foot support.

III. ALTERNATE REQUIREMENTS Alternately it is requested that an automated examination from the inside diameter, in conjunction with the vessel examination, be accepted in lieu of the surface examination from the O.D. This relief request was granted to permit this alternative for the second interval based upon examinations from the I.D. of a Surry calibration block which demonstrated sensitivity adequate to resolve a 5%

notch on the O.D. In addition, this ultrasonic technique demonstrated detection of a flaw in a mock-up which was estimated to be eighty percent of the critical flaw as described in 1980 Edition, Winter 1980 Addenda of ASME Section XI, IWB-3000 Acceptance Standards for Flaw Indications .

S2-002I3 Rev. 0 2-6 12/93

  • RELIEF REQUEST SR-001 (CON 1 T)

The demonstration of the above ultrasonic technique was witnessed by Authorized Nuclear Inservice Inspector at the Westinghouse Waltz Mill Calibration Facility and found to be acceptable. This was also demonstrated to the NRC staff on June 24, 1986 (reference NRC Letter serial #86-759, dated Nov. 12, 1986) .

  • S2-002I3 2-7 Rev. 0 12/93
  • I.

RELIEF REQUEST SR-002 IDENTIFICATION OF COMPONENTS Nozzle Inner Radius Sections (Primary Side Steam Generator)

Mark # Component # Drawing # Class 1-0lANIR 2-RC-E-lA 11548-WMKS-RC-E-lA.2 1 1-0lBNIR 2-RC-E-lA 11548-WMKS-RC-E-lA.2 1 1-02ANIR 2-RC-E-lB 11548-WMKS-RC-E-lB.2 1 1-02BNIR 2-RC-E-lB 11548-WMKS-RC-E-lB.2 1 1-03ANIR 2-RC-E-lC 11548-WMKS-RC-E-lC.2 1 l-03BNIR 2-RC-E-lC 11548-WMKS-RC-E-lC.2 1 II. IMPRACTICAL CODE REQUIREMENTS Category B-D, Item No. B3.140 requires the volumetric examination of the nozzle inside radius sections of the steam generator primary nozzles.

III. BASIS FOR RELIEF The only viable ultrasonic technique currently available to examine nozzle inner radii involves the fabrication of calibration blocks that closely simulate the O.D. and I.D .

nozzle geometry. This is necessary so that search units can be produced that will interrogate the inner radius section at precise angles. When the beam angle varies it is not possible to locate indications or discriminate between flaws and geometry. Also, in order to obtain meaningful results, the nozzle material grain structure must be such that a relatively high signal-to-noise ratio can be obtained over the required metal path distance. Additionally, for nozzles with a complex O.D. profile, examination personnel need training on the proper placement and manipulation of the search unit.

Virginia Power has previously assessed the feasibility of performing examinations on the North Anna Unit 2 cast primary nozzle inner radii. Virginia Power performed examinations on a Westinghouse Model 44 channel head which was used to train welders for the North Anna Unit 1 steam generator replacement.

This channel head is made from the same cast material (ASTM 216-WGG) as Model 51 generators which are currently installed in Surry Unit 2. We believe that the general surface profile and acoustic properties are representative of the Surry Unit 2 steam generators .

  • S2-002I3 2-8 Rev. 0 12/93
  • RELIEF REQUEST SR-002 EXAMINATION RESULTS
1. Comparison of Material Noise (CON'T.)

Figures A and Bon Attachment 1 depict the respective responses from notches 4 and 6 from calibration block VPSGINRl. This calibration block was manufactured to examine the North Anna Unit 1 replacement steam generator forged carbon steel primary nozzle inner radii. Both notch responses exhibit a high signal to noise ratio with little evidence of material noise. Figure A on Attachment 2 depicts the material noise from the Model 44 primary nozzle at the same sensitivity level. At this sensitivity level, there is no evidence of clad roll.

The first indication of sound penetration (Attachment 1, Figure B) appeared at 12db above the reference level when evidence of clad roll was detected.

2. Surface Profile The O.D. surface of the nozzle had an irregular contour which is typical of large cast products. Due to the surface contour, it was necessary to apply a large amount of couplant to maintain contact with the examination surface. The irregular surface and large amount of couplant required caused an apparent change in the sound beam angle from point to point over the examination surface. Therefore, it could not be determined where the sound beam was directed with respect to the inner radius.

CONCLUSION The Surry Unit 2 steam generator primary nozzle inner radii were not designed for ultrasonic examination from the O.D.

The nozzles are integrally cast into the channel head.

Therefore, the nozzles contain examination limitations such as an irregular O.D. profile, a rough surface condition, and an attenuating grain structure. The irregular surface causes the beam angle to change from point to point around the nozzle. The varying beam angle combined with a relatively low signal-to-noise ratio makes evaluation of the results extremely difficult. Furthermore, it would be unduly difficult to design a practical search unit for this configuration. As a result of the above limitations, it is our opinion that a full scale cast mock-up of the nozzle would be necessary to develop an inner radius examination technique, and to correspondingly provide appropriate training for the examination personnel. As such, the Code prescribed examination is deemed impractical .

  • S2-002I3 2-9 Rev. 0 12/93
  • IV.

RELIEF REQUEST SR-002 (CON 1 T.)

ALTERNATE PROVISIONS As an alternative, the areas will be visually (VT-1) examined from the nozzle I.D. using direct or remote methods per the schedule shown in Table IWB-2412-1 *

  • S2-002I3 2-10 Rev. 0 12/93
  • - ti" P(ii**.*c;*., .,::,-,1'-'~;;.:*:~~=--~=iit *~:

ATTACHMENT 1

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    • -~ FIGURE A
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Response from Notch 4 in IEL C. * -::* 531 in/us O .5-o .oo M-.. .IF'F. VPSGINRl at 9.05 11 Metal "HlCf:. =.SS O .000 in PULSER LOt! -JOg* Path distance and 79.1 dB.

  • I.I.&*:-':"'"; t:Wt.£:a t.:-~W c-utl WH.10::

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  • 1 dB FIGURE B Response from Notch 6 in
  • . VPSGINRl at 13.27" Metal Path distance and 79.1 dB .
  • S2-002I3 2-11 Rev. 0 12/93

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ATTACHMENT 2 FIGURE A Material Noise from cast nozzle at 79.1 dB.

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1 .096 in/[HIJ FIGURE B First evidence of clad roll I) I' from I.D. at inner radius.

Sensitivity is a 91.1 dB.

General noise is 20% of full screen.

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  • S2-002I3 2-12 Rev. 0 12/93
  • I.

RELIEF REQUEST SR-003 IDENTIFICATION OF COMPONENTS Nozzle Inner Radius Section (Pressurizer surge nozzle)

Weld# Component# Drawing# Class 15NIR 2-RC-E-2 11548-WMKS-RC-E-2 1 II. IMPRACTICAL CODE REQUIREMENTS Category B-D, Item BJ.120 requires the volumetric examination of the nozzle inside radius section of the pressurizer surge nozzle.

III. BASIS FOR RELIEF The Surry Unit 2 pressurizer surge line nozzle is integrally cast into the bottom pressurizer head. The nozzle is located under the pressurizer skirt and is surrounded by 78 heater penetrations. Interference from the heater penetrations and heater cables, as well as the location of the nozzle under the pressurizer skirt restricts the access to the nozzle.

This limits the examiners ability to manipulate the search unit to examine the nozzle inner radius.

The only viable ultrasonic technique currently available to examine nozzle inner radii involves the fabrication of calibration blocks that closely simulate the O.D. and I.D.

nozzle geometry. This is necessary so that search units can be produced that will interrogate the inner radius section at precise angles. Also, in order to obtain meaningful results, the nozzle material grain structure must be such that an adequate signal-to-noise ratio can be obtained over a long metal path distance.

Integrally cast nozzles contain limitations such as an irregular O.D. profile, a rough surface condition, and an attenuating grain structure. The irregular surface condition causes the beam angle to vary from point to point around the nozzle. The attenuating grain structure results in a low signal-to-noise ratio at the nozzle inner radius. Limited access to the nozzle as well as the limitations imposed by the material conditions, area dose rates and the complicated nature of the examination technique would make evaluation of any indications very difficult.

Any examination on this nozzle could only be described as "best effort", and not commensurate with the anticipated

  • exposure to perform this examination .

S2-002I3 Rev. 0 2-13 12/93

  • RELIEF REQUEST SR-003 (CON'T)

It is estimated that at least 3.675 man-rem would be required to perform this inspection and if the cables to the heater penetrations require removal to provide better access, then greater than 9 man-rem would be required.

IV. ALTERNATE REQUIREMENTS A visual (VT-2) examination of the pressurizer surge line nozzle area will be performed during the normally scheduled pressure test (Class 1) each refueling .

  • S2-002I3 2-14 Rev. 0 12/93
  • I.

RELIEF REQUEST SR-004 IDENTIFICATION OF COMPONENTS systems: outside Recirculation Spray (RS) and Safety Injection (SI)

Components: Pump casing welds .identified below Component Weld Drawing# Class 2-RS-P-2A 2-01 11548-WMKS-RS-P-2A 2 2-RS-P-2A 2-02 11548-WMKS-RS-P-2A 2 2-RS-P-2A 2-03 11548-WMKS-RS-P-2A 2 2-RS-P-2A 2-04 11548-WMKS-RS-P-2A 2 2-RS-P-2B 2-01 11548-WMKS-RS-P-2B 2 2-RS-P-2B 2-02 11548-WMKS-RS-P-2B 2 2-RS-P-2B 2-03 11548-WMKS-RS-P-2B 2 2-RS-P-2B 2-04 11548-WMKS-RS-P-2B 2 2-SI-P-lA 2-01 11548-WMKS-SI-P-lA 2 2-SI-P-lA 2-02 11548-WMKS-SI-P-lA 2 2-SI-P-lA 2-03 11548-WMKS-SI-P-lA 2 2-SI-P-lA 2-04 11548-WMKS-SI-P-lA 2 2-SI-P-lB 2-01 11548-WMKS-SI-P-lB 2 2-SI-P-lB 2-02 11548-WMKS-SI-P-lB 2 2-SI-P-lB 2-03 11548-WMKS-SI-P-lB 2 2-SI-P-lB 2-04 11548-WMKS-SI-P-lB 2 II. IMPRACTICA~ CODE REQUIREMENTS Category C-G, Item C6.10, Pump Casing Welds, requires that a surface examination be performed on 100% of the welds each interval. The examination can be limited to one pump in the case of multiple pumps of similar design, size, function, and service in a system.

III. BASIS FOR RELIEF These pumps are vertical, two-stage, centrifugal pumps, with an extended shaft and casing to allow suction from the containment sump. The motor and mechanical seals of the pumps are located at approximately the 12 foot elevation and the bottom of the casing is located at approximately the -30 foot elevation. The welds identified are at. the bottom of the pump casing, and are embedded within the concrete building structure. This makes the welds inaccessible from the outside. The small diameter of the casing (24 inch. O.D.)

and the pump shaft prevent examination from the inside diameter .

  • S2-002I3 2-15 Rev. 0 12/93
  • IV.

RELIEF REQUEST SR-004 (CON 1 T)

ALTERNATE REQUIREMENTS A visual examination (VT-1) of the accessible portions of the I.D. of the pump casing welds will be performed only if the pump is disassembled and the pump shaft

.removed for maintenance .

  • S2-002I3 2-16 Rev. 0 12/93
  • I.

RELIEF REQUEST SR-005 IDENTIFICATION OF COMPONENTS Ultrasonic calibration blocks II. IMPRACTICAL CODE REQUIREMENTS ASME Section XI, Appendix I specifies requirements for ultrasonic calibration blocks. Specifically, calibration blocks must meet the requirements of ASME Section V (vessels

> 2" thick) or ASME Section XI, Appendix III (piping and vessels~ 2 11 ) as supplemented by Table I-2000-1. The existing calibration blocks at Surry Unit #2 were designed and fabricated before the guidelines of ASME Section XI were developed and approved.

The blocks for piping and vessels~ 2" do not meet the recommended design specified by Figure III-3230-2 for thicknesses< 1 11 in that the notches are not staggered. Also, the notches in most of the piping blocks are located 1* "t" from the end of the block instead of 1 1/2" as specified by Figure III-3230-1. The vessel calibration blocks used for the RV head-to-flange weld, SG primary side tubesheet-to-head weld, and pressurizer welds are partially clad instead of fully clad as shown by Figure T-441.1 of ASME Section V, Article 4.

III. BASIS FOR RELIEF Meeting the above new ASME Section XI requirements would require the fabrication of new calibration blocks.

Satisfactory ultrasonic system calibration can be performed with the existing calibration blocks. Use of the existing calibration blocks also allows correlation of ultrasonic data from previous interval examinations as required by IWA-1400(h). The location of the notches in the piping calibration blocks provides adequate signal separation for sweep calibration. Distance-amplitude calibration down to the clad-to-base metal interface, as delineated by Nonmandatory Appendix B to Section V, Article 4, can be performed from the unclad portion of the clad side of the existing vessel calibration blocks.

IV. ALTERNATIVE REQUIREMENTS It is proposed that the existing calibration blocks be used during the third inspection interval .

  • S2-002I3 2-17 Rev. 0 12/93
  • I.

RELIEF REQUEST SR-006 IDENTIFICATION OF COMPONENTS ISI class 1 and 2 piping, vessel and component welds II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code. 1989 Edition, IWA-2600, weld reference system III. BASIS FOR RELIEF The original construction code used at Surry Power Station, ASA B31.1-1955 with the applicable nuclear code cases, did not establish a weld reference system. Immediate establishment of a weld reference system cannot be practically attained within the scope and schedule of existing outages.

IV. ALTERNATE REQUIREMENTS Surry Unit 2 has recently updated its weld isometrics, providing a detailed identification of weld location. These drawings will be used in tracking and locating welds .

In addition at the time welds are examined volumetrically for program requirements, a reference will be established for each weld, indicating a zero point and direction of examination. Welds which contain recordable indications (RI) shall be marked to ensure location of the indication, using appropriate reference marks. This reference system and marks will be permanently fixed on the weld .

  • S2-002I3 2-18 Rev. 0 12/93
  • I.

RELIEF REQUEST SR-007 IDENTIFICATION OF COMPONENTS Pressure retaining welds in the reactor vessel and vessel nozzle area examined by the automated vessel tool inspection device.

II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, IWA-2600, weld reference system III. BASIS FOR RELIEF The automated tool establishes its reference point using an existing zero reference in the reactor vessel. This point allows the device to repeat examination locations without the necessity of any other reference systems. It accomplishes this by the use of an electronic encoder system which provides for sufficient repeatability.

IV. ALTERNATE REQUIREMENTS The automated vessel tool examinations will continue to establish it's reference system based upon the existing zero reference. No other system is planned or deemed necessary .

  • S2-002I3 2-19 Rev. 0 12/93
  • I.

RELIEF REQUEST SR-008 IDENTIFICATION OF COMPONENTS ISI Class 1 Pressure Retaining Welds in Piping II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, requires that notes l(b) and 2 of Category B-J, table IWB-2500-1 be used in the selection of welds for examination.

III. BASIS FOR RELIEF The second interval selection was based upon the 1974 Edition with Summer 1975 Addenda (74/S75) of ASME Section XI. As a result notes l(b) and 2 cannot be applied without some programmatic additions and modifications. In addition, although stress and utilization calculations exist for Surry Unit 2, no correlation exists with actual weld locations.

Total reuse of the second interval plan is not desirable, since even though the 74/S75 requirements were met, distribution of welds selected did not equitably cover certain line functions and designs .

IV. ALTERNATE REQUIREMENTS ISI Class 1 piping welds will be selected for examination such that 25% of the total number of welds are examined during the interval. The 25% sampling will include terminal ends as they appear on our plant isometrics as no corresponding stress calculations exist (ref. NUREG-0800, BTP 3.6.2-13). Terminal ends will include extremities of piping runs connected to vessels and pumps. Pipe integral attachments that act as rigid constraints to piping motion and thermal expansion shall have the first weld upstream and downstream of the integral attachment selected. In piping runs which are maintained pressurized during normal plant conditions for only a portion of the run, the first normally closed valve is a terminal end. The weld on the high pressure side of the valve will be included in the interval selection. Additionally all branch connections will be selected. Terminal end welds are identified in the third interval plan by the designation "Terminal End" in the comments column. The welds selected will be evenly distributed based upon line size, line function, and line design to the extent practicable. These selected welds will be examined in future successive inspection intervals to the extent allowed by code editions approved at that time .

S2-002I3 Rev. 0 2-20 12/93

  • I.

RELIEF REQUEST SR-009 IDENTIFICATION OF COMPONENTS Class 1, 2, and 3 Integrally Welded Attachments II. IMPRACTICAL CODE REQUIREMENTS Examination Categories B-H, B-K-1, c-c, D-A, D-B, and D-C of ASME Section XI, 1989 Edition, with regard to Integrally Welded Attachments.

III. BASIS FOR RELIEF Code Case N-509, Alternate Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments,Section XI, Division 1, is not currently approved by Regulatory Guide 1.147 for use. 10 CFR 50.55a footnote 6 notes that the use of other Code Cases may be authorized by the Director of the Office of Nuclear Reactor Regulation upon request pursuant to 10 CFR 50. 55a ( a) ( 3) . As.

such, Code Case N-509 is requested for use on Surry Unit 2 in the third inspection interval.

The current Code requires a certain size base material design thickness before examination is required on Class 1 or 2 integrally welded attachments. This size limitation is apparently arbitrary with not technical basis. The current Code also has no inspection requirements for Class 1 integrally welded attachments for piping, pumps, and valves (B-K-1) for the third inspection interval of Inspection Program B (Surry's). Additionally, there is no selection criteria for Class 3 nonexempt integrally welded attachments, requiring 100% examination. These deficiencies have been corrected in Code Case N-509.

IV. ALTERNATE REQUIREMENTS Code Case N-509 will be used in its entirety. Code Case references to the 90 addendum of ASME Section XI are the same as the provisions in Code Case N-491, which has been implemented by our support program .

  • S2-002I3 2-21 Rev. 0 12/93
  • I.

System:

RELIEF REQUEST SR-010 IDENTIFICATION OF COMPONENTS Chemical and Volume Control (CH)

Component: Regenerative Heat Exchanger (2-CH-E-3)

Welds/Components Description Class 1-06 nozzle-to-vessel 1 NIR-06 nozzle inside radius 1 1-08 nozzle-to-vessel 1 NIR-08 nozzle inside radius 1 1-09 nozzle-to-vessel 1 NIR-09 nozzle inside radius 1 1-11 nozzle-to-vessel 1 NIR-11 nozzle inside radius 1 1-13 nozzle-to-vessel 1 NIR-13 nozzle inside radius 1 1-15 nozzle-to-vessel 1 NIR-15 nozzle inside radius 1 II. IMPRACTICAL CODE REQUIREMENTS

  • III.

Examination Category B-D (Inspection Program B) requires in item numbers B3-150 and B3.160 a volumetric examination of the nozzle-to-vessel weld and nozzle inside radius section.

BASIS FOR RELIEF The joint design of the above nozzle welds specifies a 3 11 schedule 160 weldolet jointed to a 9.25" O.D. x .875 11 thick vessel. These welds were not designed to be volumetrically examined from the outside diameter. The configuration of the weldolet precludes axial ultrasonic examination from the nozzle side and circumferential examination in either direction. This limits volumetric examination to a single axial scan from the vessel side of the nozzle. It is our opinion that a meaningful ultrasonic examination cannot be performed on the weld or inner radius with a single axial scan from the vessel side. This is due to the small diameter of the vessel and weldolet, the change in dihedral around the joint results in a corresponding change in the ultrasonic beam angle. This makes position measurements unreliable. It would also be necessary to extend the beam path to at least two full Vee paths, which will further complidate this examination. These limitations would substantially limit our ability to discriminate flaw indications from geometry existing around the joint .

S2-002I3 Rev. 0 2-22 12/93

  • RELIEF REQUEST SR-010 (CON 1 T)

The configuration also precludes placement of film on the outside diameter for radiography, and the inside surfaces are inaccessible. It is our opinion that the gain in assurance of component integrity from this limited examination is not commensurate with the anticipated dose expenditure. Figures 1, 2, and 3 are provided to support this request.

IV. ALTERNATE REQUIREMENTS Th~ outside surface of the nozzle-to-vessel welds shall receive a surface (liquid penetrant) examination. In addition a visual (VT-2) examination shall be performed in conjunction with the pressure testing required by IWB-5000 and/or Code Case N-498 .

  • S2-00213 2-23 Rev. 0 12/93

3 2 RECiENERATIYE ~

A HEAT EXCHANGER ro f--'

.-. 2-CH* I-'*

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  • I.

Weld#

RELIEF REQUEST SR-011 IDENTIFICATION OF COMPONENTS Component ID# Description Class 1-07 2-RC-E-2 shell to head 1 1-02 2-RC-E-2 longitudinal 1 Drawing: 11548-WMKS-RC-E-002 II. IMPRACTICAL CODE REQUIREMENTS The 1989 Edition of ASME Section XI, Category B-B, items B2.11 and B2.12 requires a volumetric examination of 100% of the circumferential shell to head weld and 1 foot of the intersecting longitudinal weld.

III. BASIS FOR RELIEF The pressurizer is covered with an insulation support ring (attachment 1). The insulation support ring is 6 inches wide, where it interferes with the examination. This ring prevents complete volumetric coverage of both the upper shell

  • to head weld and intersecting longitudinal weld. Total removal of the insulation support ring is considered.

impractical due to high anticipated exposure levels estimated at 18.4 man-rem. Partial removal of the support ring at the mechanical connection would allow some increase in coverage near the mechanical connection, where the support ring could be spread apart. However, the actual area of weld made accessible to this increased coverage is estimated to be very small in relation to the overall weld length, because the insulation support structure is rigid, interconnected with cross supports, and welded to the supports for the safety valves and power operated relief valves. The intersection of the circumferential shell to head weld and longitudinal welds is physically located behind one of these supports.

Examination coverage of this area would not be improved by partial removal at the mechanical connection.

Any removal of the mechanical connection and spreading apart of the support structure would increase the risk of misalignment problems, and warping of the structure. This risk coupled with the marginal increase in examination coverage, makes partial removal of the insulation support structure also impractical .

  • S2-002I3 2-24 Rev. 0 12/93
  • IV.

RELIEF REQUEST SR-011 (CON 1 T)

ALTERNATE REQUIREMENTS A volumetric examination will be performed to the extent practical on welds 1-07 and 1-02. No extended beam path examinations can performed, since the pressurizer is a clad vessel. The estimated coverage in the perpendicular beam path directions are detailed in attachment 2. The parallel beam path estimates are as follows:

Weld# 450 (7&8) 6Qo (7&8) 1-07 26% 26%

1-02* 100% 100%

  • (accessible length only - approximately 7.5")
  • S2-002I3 2-25 Rev. 0 12/93

Reliet Request SR-011, Attachment 1, Page 1 of 4 ATTACHKEH'l' 1

Kelie! keguest Si;S I/ I I I Art:acnment: .L, .r'age .:. 0~

PAGE I OF" 2 ,

- - .I

-~IIISLILATIOII

. , '? SUPPORT

.* ( F"AMt IMSULATIOM IUf'f'O"T AINGi.S RING

./

CONNLC110N wELD*a - - PRtSiUIIUZE"

&tN*ND su,,.o~'T W[LDtD CT,.I')

IN5ULATION 5UPPORT 150

,roR. STUDY OF INTtRFf/UNClS WITH I~ I NOl o, rRt.S~UltlZER wt'-05

  • 7 t 2 )

SURRY POWER STATION PRESSURIZER WELD UT INTERFERENCE UNIT 2 DRAWN BT: EH.G~: ~TE:

"'" ~~~ l()*;)_0-5k

PAG ......

_ _OF'_2 2 f .JU. C 1.10. Z J. -

i 1NSULATIOM j IU,,OAT 1 fUNG,-

'-- -flOlf..V su,PO"-T

,s*---r **

WU,0 IN IP'&CTIOH A~£A SE..C.TIOt-J A*A SUP,.OIIT AING,

~lt.t!!IU~l'ZE~

SECTJON 5- B SURRY POWER STATION PRESSURIZER WELD.

UT INTERFERENCE UNIT 2

0)-----

TYP(SEE NOTE 5) f LOCATION PLAN

'- *f NOTES:

1. SCALE, NONE
2. STUDS SHALi:. BE PRELOADED TO PRODUCE A MEASURED STRAIN OF.028"!,002"1 THIS TO BE DONE WHILE THE 9 DISC SPRINGS FACING INWARD PLAN FOLLOWING CONDITIONS EXIST:

PRESSURIZER SHELL AT AMBIENT TEMP& ATMOS.

9 DISC SPRINGS FACING OUTWARD PRESSURE AND STRAPAT AMBIENT TEMP! 15*F

3. INSULATION ON 9.JPR:JRT COLUMNS TO BE .

5" +~ IN ACCORDANCE WI DC 69 1a ¢ BORE_o* AT FINAL.INSTALLATION THIS DMEN~ION TO BE 1~!t-4.

5- 5PECIFIC WELDING & NOT PROCEDURES SHALL BE DEVELDPED 2

FOR THIS JOINT IN ACCORDANCE WITH SITE SPECIFIC EQUIPMENT AND CONDITIONS

5. SEE SK.738955/8471-M-40 U) 7.

a.

REFERENCE DWG. 11448-FP-9D-7 DESIGN BASED~ CALC.tO 14937.03-MN(B)-263-JC REV. O

,-1_ - ,- - - - - -.. __,~,

,_ - - - - - t---+--+--- 9. DISCARD "!JOINT SPACER' (SEJ:: REF. DWG.FINAL DRAFT

'-L- - - - - - - _t---f.-- 10. SEE SK. 738955/8471 -M- 41143 & 45 .

Q. A. CAT. I SWEC 14937.03 DC 71 tAJCLF AR SAFE TY REL.A:fED NU TS AND SPRING WASHERS

,NOT SHOWN FOR CLARI TY . II 4

Tl1£ INfOftlllATION OH THIS DRAWING MAl NOT It(

COPIED Olt US[C roR 'JTHER 1"Aff *h[ *::ONSTRUCTIOH 111.AINTENAHCE *JR R(PIIIA Of THE PLANT FACll 11 l tv'l'frDdlrn ,.., "f.,r TITI C hi ,v-v

Relief Request SR-011, Attachment 2, Page 1 of 3 A'll'ACBKEHT 2

Surry Unit 2 Pressurizer Veld 7 Por-tlal ExoMlna-tlon [voluotlon

<Ha. tched oreos represent no e><oMlnatlon coverage>

V*ld 7 V*ld 7 2 Scon S Scan 4:,* tr.!* Transduc*r 45" 1,2* Tn,n9duc*r Vessel Side Vassel Side JB.53 Sq. In. E><Gl'I. Voh.rie 3B,:53 Sq. In. [,cal'I. Volul'le 24.93 Sq. In. not e><ol"lln!'d 13.70 Sq. In. E><or11ned R*c:µr*d [>c6"1natl0n Boundary A-1-C-D 3e.e9 Sq. In. no't e,cal'llned R*quir*d [1Co111not1011 ICLrtdory A-1-C-D 6.24 Sq. Jn. [><O.rtlned 36Y. E><ol'llned 16:t. E><cu,lned Vrld 7 V*ld 7 S Scan 2 Scan

,o* 112* Transduc*r 80' J/1!' Transducrr Heo.d Side Vessel Side Vessel Side D

38.53 Sq. (rL E><or1. Volur1e R1QU1r1d [lt'CINliGtlan loundary 38.53 Sq. In. E><oM. Volur,a 20.47 Sq. In. no.t: e><o.l'llned A-1-C-D 27,91 Sq. In. not e><ol'lln!'d Rrqulrrd [11a""'atl0n loundory 18.06 Sq. In. E><aMlned 10.62. Sq. In t><nrt1ned A-11-C-D

,47;: E,co.rtlned 281. E><o.nln~d

  • .~------.J*

Surry Unit 2 Pressurizer

'w'eld 2 Partial ExaMlnatlon Evalua tlon

<Hatched oreos represent no exoM~otlon coverage>

I v.1d e 2 Sca.n without s...,part Ring lntrrfrrrnc

/2' Tronsduc,r p V s.c,pa,.t I

I

-- lu"'*~ ... l B A ---- lupporl(a. . -

I I

\ I 3B.53 Sq. In. Exo.M, VoluM~ \

I I I 3B.53 Sq. In. E><o.l'I. Volul"le 0.0 Sq. In. not exal"llned \ I 0.0 Sq. Jn. not exo."ned 38.53 Sq. In. Exo.Mlned \ I / 3B.53 Sq. In. E><o.l'!lned IOOi: Exar1lned

\I) ' ,

100:.: E><o.r1tned D C A*"-"r*d [1u1.111nat1on Boundary A-1-C-D TTP.

Count~rclockwls~

Clockwl5e V*ld 2 2 Sc4n without Support Ring mt*rhrl!nct 60' 1/2' Tnin1duc1r

~v lul'f""'t J I I JI I 38.53 Sq. Jn. Exo.M, VoluMe I I 38,53 Sq. Jn, Exo.r,, Volur,e \

0.0 Sq. In. not exaruned I I 0,0 Sq. In. not exoMlned 'I 38.SJ Sq. In. Exal'llned I 38.53 Sq. Jn. Exo.r,lned '\

IOOio: Exo.r,1ned I 1007. ExoMln~d ......

', ,,~ D C D C NOTE*

'w'eld 2 coverage estlno.te o.pptles only io the area of the weld not covered and r,o.de Inaccessible by the Insulation suppori ring.

Of the Code required exo.r,1nat1on area. the lnsula tlon support ring covers the iop 4 lni;hes of' weld 2, o.llowlng o.ppro><IMo. tely 8 Inches to be o.ccess;lble.

  • I.

RELIEF REQUEST SR-012 IDENTIFICATION OF COMPONENTS Class 1 and 2 longitudinal welds in piping II. IMPRACTICAL CODE REQUIREMENTS Table IWB-2500-1, Examination Category B-J, items B9.12 and B9.22, and Table IWC-2500-1, Examination Category C-F-1, items C5.12, C5.11, and C5.42, and Examination C-F-2, items C5.52, C5.62, and C5.82 pertaining to longitudinal welds in piping.

III. BASIS FOR RELIEF ASME Section XI in Code Case N-524 provides alternative examination requirements for longitudinal welds in piping.

This Code Case is not currently approved for use in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability ASME Section XI Division 1. Footnote 6 to 10 CFR 50.55a states that the use of other Code Cases may be

  • authorized by the Director of the Office of Nuclear Reactor Regulation upon request pursuant to 10 CFR 50.55a (a) (3),

which requires that the proposed alternative provide an acceptable level of quality and safety, or that compliance with the existing Code requirements would result in hardship or unusual difficulty without a compensating increase in level of quality and safety.

Code Case N-524 provides an acceptable level of safety and quality by concentrating the examination in the high risk location associated with the intersection of the circumferential and longitudinal weld in piping.

Additionally the Code Case eliminates the portion of the current Code examination volume requirements beyond the intersection area. This area is perceived by the ASME Code as being relatively low risk. Continued examination of this area would significantly increase the exposure of examination personnel with a minimal increase to safety.

  • IV. ALTERNATE REQUIREMENTS The requirements of Code Case N-524, Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping Section XI Division 1, shall be followed .
  • S2-002I3 2-26 Rev. 0 12/93
  • I.

Relief Request SR-013 IDENTIFICATION OF COMPONENTS system: Residual Heat Removal Components: Integral Attachments on Drawings 11548-WMKS-RH-E-lA, and 11548-WMKS-RH-E-lB as listed below; Component Integral Attachment Class 2-RH-E-lA HOOl-1 2 2-RH-E-lA H002-l 2 2-RH-E-lB HOOl-1 2 2-RH-E-lB H002-1 2 II. IMPRACTICAL CODE REQUIREMENTS Category c-c of the 1980 Edition, Winter 1980 Addendum of ASME Section XI, item CJ.10 requires a surface examination on 100%

of the required areas of each integral attachment. In case of multiple vessels of similar design and service, the required examinations may be conducted on only one vessel. Where multiple vessels are provided with a number of similar

  • III.

attachments, the examination of attachments may be distributed among the vessels.

Basis For Relief These heat exchangers and integral attachments were designed and constructed to ASME Section VIII, 1965 Edition, Winter 1966 Addendum. This Code limited the examination of the integral attachments to a visual type exam requiring (UW-38) that visible defects, such as cracks, pinholes, and incomplete fusion and defects detected by the hydrostatic test be removed.

These fillet weld integral attachments never received nor were they required to receive a surface examination under the Construction Code. Additionally the initial preservice examinations at Surry were limited to Class 1 components, since Class 2 and 3 had not yet been placed into the Code, as such no preservice surface examination was conducted on these integral attachments. The actual surface condition is indicative of a visual only type exam from this time period .

  • S2-002I3 2-27 Rev. 0 12/93
  • Relief Request SR-013 (CON'T)

Attempts to examine these attachments as part of the present inservice inspection program (surface exam) have resulted in reports which identify the general rough surface condition.

The most recent attempt was rejected by the Level III, as unacceptable to perform the surface examination due to the rough surface condition. An evaluation of the man-hours and dose requirements necessary to properly prepare the surface, which will include both grinding and welding of the integral attachments, has been completed (attachment 1). This estimate calculates an expenditure of 11.74 man-rem, if all attachments are properly prepared for a Section XI preservice examination.

This dose expenditure and outage maintenance addition is considered impractical considering that the original Construction Code accepted the surface condition visually.

IV. ALTERNATE REQUIREMENTS The integral attachments shall be inspected visually (VT-3) for cracking or other conditions described as unacceptable in the original Construction Code. This alternative was approved for the Surry Unit 2 Interval 2 Program by letter #93-026 dated 1/12/93 .

  • S2-002I3 2-28 Rev. 0 12/93

VIRGINIA POWER MEMORANDUM To: Mr. D. L. Rogers Office/Location ISi Engineering From: M. S. Whitt Office/Location Civil Design Date December 20, 1991 RHR HEAT EXCHANGER SUPPORT WELD REPAIRS The purpose of this memorandum is to provide man-rem estimates for repair of the welds on the RHR heat exchanger support lugs for both heat exchan.gers in Unit 1 and Unit 2, as requested by your memorandum of October 3, 1991.

During the Unit 2 Refueling Outage, DR #S1-91-0582 was written due to the NOE failure of an inteornllv attached support lug on 2-RH-E--1 A. During the subsequent evu:udtion, which tnc!uood visual examinatiora~ of supports on bot~ 2--RH-F.-1 A r:1no 1 B, it was determmt:;'1 that the welds in question had never been prepared to allow for a surface NOE exam as required by ASME Section XI, because the heat exchangers were originally fabricated to ASME Section VIII.

Your memorandum of October 3, 1991 requested that Design Engineering work with NSS Construction and the ALARA coordinator to develop comprehensive and accurate man-hour and radiation exposure dose estimates for weld repairs to provide acceptable welds to allow NOE inspections of the welds.

By memorandum of October 16, Engineering requested NSS Construction to develop the required man-hour estimates. These estimates were provided in the memorandum from Mr. W. D. Corbin of December 10, 1991 (Attachment 1). These man-hour estimates were used by the ALARA Coordinator to estimate man-rem exposure (Attachment 2). The total exposure for this work is calculated to be 11. 74 man-rem for each unit, and 23.48 man-rem for both units .

Mr. D. L. Rogers

  • December 20, 1991 Page 2 If you have any questions or require further information, please contact me.

M. S. Whitt cc: T. B. Sowers A. R. Fletcher T. R. Huber E. W. Throckmorton - IN3W W. D. Grady M. A. Ringler L. J. C1_1lliv~ri

  • T. F. S*i.eed - l\L.A.R.A Cnordinatc~

C. E. Sorrell - IN 1 NW DEO File

  • ATTACHMENT 1 The following man-hour estimates were calculated for the weld repair anc.i inspection work for the RHR heat exchanger supports .

t -* Memorandum

  • DI-HMD

,.,,,,,,CAIIO' . . ~

To Frcm Distribution

w. D. Corbin
.' .. .

' ~

' . . ~

Innsbrook Technical Center December 10, 1991 COIIPLBTBD BSTIIIATB HUCLBAll BNGIRBBllIRG SERVICES SJJBIX PQRQ SB'1'IQN Estimate no. SY 18 SAO TYPE or ESTIIIA'l'B PROJECT TYPE

_x_ STUDY IRi CONCEPTUAL DCP DEFINITIVE EWR BID OTHER ALARA M-H EST, Attached is an estimate for RHR BX SUPPORT WELP REPAIR MANHOURS *

  • ThiE e~timate is being provided for the followi~g purpcs~:

Basis for Project Authorization Budget information Final controls packa9e Cost benefit analysis or study Other (ALARA man-hour estimate)

Please review and provide your comments. The Estimator is Mr. I.

West (ext. 3444). Corrections or revisions needed during the review process, however, should be bro~ght to the attention of Bill Miller on ext. 2553.

w. D. Corbin ATTACHMENT cc: Mr. H. H. Blake, Surry Site Services Mr. R. M. Cramer, Surry Site Services Mr. D. w. Buster, Surry Site Services

~-~ .. K:.1-' ha. Nbi~..,a.. SUg,ry~'DBO' *. ,-:.

  • Mr. T. F. Steed, Surry ALARA Coordinator Mr. M.A. Ringler, Surry Materials Engineering Estimating File - IN3NW
  • SURRY POWER STATION NUCLEAR ENGINEERING SERVICES BSTIKATB NO. SY 18 SAO SURRY RBR SUPPORT WELD REPAIRS SmuwtY DATA SOURCES
1. Sketches and data received from VA Power - Surry DEO is the basis for the design evaluated in this estimate.

This information was contained in an estimate request memorandum from Mr. M. s. Whitt to Mr. H~ H. Blake, dated October 16, 1991.

2. The information and scope of work was further defined through discussions with Mr. Mike Ringler, who is familiar with the condition of the existing welds.
3. Radiological conditions in the work area and in the general area were obtained from St~tion ALARA personnel *
  • l{ETBC!"OLOGX Man-hour estimates were developed for Construction, NDE, QC, Materials Engineering and HP personnel who would be involved in the welding repair of RHR Pump supports. Composite labor costs are furnished for VA Power and NSS personnel.

SCOPE OF WORK The scope of work covered by this estimate is based on weld repair of eight (8) RHR heat exchanger pump supports as per Attachment 1 of the above referenced memorandum (copy attached).

ASSUMPTIONS AND OUALIFICATIONS

1) It was assumed, through discussion with site personnel, that the weld repair would basically require one shift per support. Personnel involved in the repair work and verification would include one welder, one grinder, fire-watch, runner, Foreman, Material Engineer, NOE, QC, and BP tech.
  • 2) It was further assumed that personnel related to NOE, QC, Material Engineering and supervision would only require approximately 25% of the time of the repair effort.
  • 3) Through discussions with Station ALARA personnel, contact dose rates and dose rates within 12" of the weld area were obtained. These rates were utilized to determine the degree of difficulty and maximum stay-time for welders and grinders. It should be noted, dtie to the high radiation, that several welders and grinders may be required to sequentially complete the repairs. -
4) It was assumed that only one weld would be repaired at

..:.-, a time (including NDE and QC verification). This will allow sufficient strength in

  • the remaining welds to support the heat exchangers without temporary supports.
5) All manhours allow 15% for rework of new welds to allow for NDE dye-penetrant testing and repair of further indications.
6) The following specific qualifications apply to the weld repairs:
a. The weld& are bi-metallic (CS - SS).
b. The existing welds will not be completely removed; existing NDE indications and flaws will be repaired.
c. A bi-metallic weld build up will be performed tc dimensions spec!fi~d on Attachment 1.
  • d.

e.

f.

g.

The h~ttom horizontgl ~dld is ~naccessible, and will not be worked.

The top horizontal weld will be 5/8".

No scaffolding is required.

Insulation consists of removable blankets.

3) As per directions in the referenced memorandum, man-hours were broken down to allow both Unit 2 heat exchangers to be worked in 1993, one Unit 1 heat exchanger in 1994 and one Unit 1 heat exchanger in 1995.
8) The manhours are further broken down by- very low dose.

rate, low dose rate and high dose rate in order to facilitate ALARA dose rate calculations.

TABLE 1 - RBR PQKP SUPPORT REPAIR IWfflOQR SQIDIAllY 1993 1994 1995 U2RO UlRO UlRO ACTIVITY 2A/B CMHl lA CMH) lB (ffl) very low dosa. 180 90 90 Mob-demob & dressout Low dose in Containment 140 70 70

  • High dose at work area TOTAL ll.Q.

450 il 225 ll 225

,_______. ______________________________________________________* ***-------------------------. --------------------------*---1 I TIMS: 15:03:23 !ESTIMATE DETAIL aEPO'ltT II DATE: 4-DBC-1991 I 1----------------------------------------------------------------------**---------------------------------~~-------------------------I I ITATIOB: SURRY . UBIT: DISCIPLIBE: NECEAB1CAL EST. TYPE: STUDY ESTIMATE: SY 18 SAO I I ~aOJSCT: BRR BX SUPPORT aSPAIR W.P. DISCRIPTIOB: ESTIMATED IIY: SHEET: ~ - 1 I I PLAIIT DSSCaIPTIOR: CHECKED BY: WIST DATI ISTAII.: 4-DIC-1991 I 1----------------------------------------------------------------------

1 ACTIVITY RANI: 01 -------------------------------------------------------------

DISCRIPTIOB: RHR HEAT EXCHABGr.R SUPPORT WELD REPAIRS 1-----------------------------------------------------------------------------------------------------------------------------------

I I I I I MATERIAL I MARHOURS I LABOR I COB I SUIICOIITRACTIBG I I CONN I I I I UBIT I NATL I UNIT I ,Rot TOTAL I coKPI LABOR I EQUIP I UBIT I su11cNTRI TOTAL I CODS I 01scaIPTIOB IUNI QTY I NATL I cosT I NBHRS I PTR! NRHRS I RATEi cosT I TOTAL I SUIICRTRI cosT I cosT

!------+--------------------+--+------+---------+--------+--------+----

i--~-..111 1osssaAL Aa1A, Fias LSI 1 0.000 o 35.ooo 1.or

+-----+-------+--------+--------+--------+----------

35 24.00 140 o o.oo o 840 I IWATCB/BUBRIR/QC/BD~ I I fNAT IBG/FORNAR/BP, I INSB x **s.

5 I 20 IWORK ARIA WILDIR LSI 1 0.000 0 20.000 1.00 20 24.00 480 0 0.00 0 480 1 IGRIBDIR I I 12 NSR X 10 HRS. I 121 1woaK ARIA QC/RDS/HP LSI 1 0.000 0 12.500 1.00 13 24.00 312 0 o.oo O 312 I INAT SBG/FORNAB I I I5 NIB X2

  • 5 IJRS
  • I 100~ INOII/DINOII/DRISS OUT LSI 1 0.000 0 45.000 1.00 45 24.00 1,080 0 0.00 0 1,080

( I

  • II*c ALL PsasoaasL I I I 9 NH X 5 HS. I I I I I 1--------------------------------------------------------+---------.

!------ ACTIVITY TOTALS I 01

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i 1131 I a,,121 01 I 01 2,,12

,------ DISCIPLIH TOTALS I 01 I 1131 I 2,1121 01 I 01 2,712 1----------------------------------- .--------------------------------------------~------------------------------------------------

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  • NlR'iH CMO..INA POWER RICK<<NJ*VIAGINIA DMR RHR HEAT EXCI-IANGER SUPPORT

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~ DATE SM OF

ATTACHMENT 2

  • The following man-rem exposure estimates were calculated by the ALA RA Coordinator using the man-hour estimates provided by the memorandum from W. D. Corbin dated
  • December 10, 1992 .

AURA Evaluatlan llllmff: _ __...AI....~

DOSE TASIC PEOPLE JCI SITE JQI SITE UTEI ESTIMATED NO. TAIi'. DESCIIPTION AND LOCATION HG'D IUtATICII NAII-IIOUH (lll/llr) NAN*REN I

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  • Component(s):

RELIEF REQUEST 1 Piping and Valves located on station print 11548-CBM-087A-3:

Component Connected PipingComponent MOV-2701 14"-RH-101-1502 MOV-2700 Function Residual Heat Removal (RH)

Class Class 1 Section XI Code Requirements For Which Relief Is Requested Class 1 System Hydrostatic Test, IWB-5222, as modified by Code Case N-498.

Basis For Relief During a normal hydrostatic test of the primary, MOV-2700 is closed in addition to MOV-2701. This prevents pressurization of MOV-2701 and the piping between the two MOVs. Both valves are closed to prevent possible overpressurization of the Residual Heat Removal System.

Alternate Testing Method This area will be tested in conjunction with the normal Class 2 N-498 pressure test at the pressure required by the adjoining Class 2 system .

  • S2-002I3 2-29 Rev. 0 12/93
  • RELIEF REQUEST 2 Component(s): Open ended intake piping before the first shut-off valve in non-closed systems located on 11548-CBM-071A-3.

Function Circulating (CW) and Service Water (SW) System Class 3 Section XI Code Requirements For Which Relief Is Requested System Hydrostatic Test, IWD-5223 Basis For Relief The Code* addresses the problem of performing hydrostatic tests on open ended portions of discharge lines beyond the last shut-off valve in non-closed systems in IWD-5223(d). A similar problem exists for the intake piping at Surry Unit 2 as it is non-isolatable for the increased pressure requirements of a hydrostatic test.

Alternate Testing Method

  • As an alternative, the requirements applied to open ended portions of discharge lines (IWD-5223(d)) will be applied. In this case confirmation of adequate flow during system operation shall be acceptable in lieu of system hydrostatic test .
  • S2-002I3 2-30 Rev. 0 12/93
  • RELIEF REQUEST 3 Component(s): Piping and components located on the following print:

11548-CBM-071B-3 Function Service Water System Class 3 Section XI Code Requirements For Which Relief Is Requested System Hydrostatic Test - IWD-5223 Basis For Request The service water system on the above mentioned drawing is used for cooling component cooling water for the charging pumps and lube oil for the charging pumps. The system was designed without the use of a safety or relief valve due to the low pressure output of the charging pump service water pumps which is 50 psig (max). The only other possible pressure source for the system would occur if an extensive heat exchanger leak occurred at either 2-SW-E-lA or lB between component cooling and service water. This pressure could

  • be no more than 57 psig the maximum discharge pressure of 2-CC-P-2A or 2B. Using a design pressure (PD) of 100 psig would be excessive for this system as the maximum pressure potential is 57 psig.

Additionally this piping connects to fiberglass reinforced plastic piping, which has a design pressure of 60 psig. In some instances this piping could be included in the hydrostatic test boundary.

Alternate Testing Method As an alternative, it is requested that 60 psig be used as this systems PD value .

  • S2-002I3 2-31 Rev. 0 12/93

Component: Piping and components between the following valves located on drawing 11548-CBM-068A-3, Sh.3 of 4:

Valves Line Valves Class 2-FW-145 l-WAPD-115-601 2-FW-149 3 2-FW-146 3 2-FW-629 3 2-FW-160 1-WAPD-117-601 2-FW-164 3 2-FW-161 3 2-FW-628 3 2-FW-175 l-WAPD-120-601 2-FW-179 3 2-FW-176 3 2-FW-627 3

  • II. IMPRACTICAL CODE REQUIREMENTS ISI Class 3 System Hydrostatic Test required in IWD-5223 of ASME Section XI, where design pressure is 1432 psig and the resultant test pressure is 1576 psig.

III. BASIS FOR RELIEF Three pressure reducing orifices 2-FW-R0-200A, 2-FW-R0-200B, and 2-FW-R0-200C provide during normal operation a pressure drop from approximately 1200 psig to 110 psig. The system design takes advantage of this pressure drop by incorporating lower pressure rated piping downstream of the orifices.

However the higher pressure rated piping continues beyond the orifices for some distance. The actual pipe design pressure class change occurs at downstream check valves 2-FW-148, 2-FW-163, and 2-FW-178 and manual valves, 2-FW-146, 2-FW-161, and 2-FW-176. The check valve operation, however, does not allow separation of the lower pressure class system, when testing the higher class piping in accordance with IWD-5223.

IWD-5223 requires a test pressure of 1576 psig on the higher pressure class components in question. The downstream connecting piping has a design pressure of 150 psig and a corresponding test pressure of only 165 psig. As such, the connecting components would be overpressurized during the

  • S2-002I3 2-32 Rev. 0 12/93
  • RELIEF REQUEST 4 (CON 1 T) required Section XI test.

The test boundary could be backed up to manual isolation.

valves 2-FW-147, 2-FW-162, and 2-FW-177, however hydrostatically testing this test boundary would also pressurize the auxiliary feedwater pumps and their suction connection. These pumps have welded discharge connections and cannot be isolated from the test boundary due to the absence of a drain or vent valve in the area identified above. There is a flange connection on the suction side of each pump, however using this flange for isolation purposes is considered difficult due to the piping arrangement, and susceptibility to cold spring misalignment problems.

Typically centrifugal pumps are hydrostatically tested at a pressure based on the suction side of the pump as described in IWA-5224(d) of the Code, which prevents any potential overpressurization concerns, or the need to use pump flanges as isolation points.

The basis for relief then is two-fold. The first impracticality is the overpressurization of piping and components downstream of the pressure reducing orifice and the design pressure class rating change. The second

  • IV.

impracticality is the incorporation of the auxiliary feedwater pumps into the test boundary due to the lack of vent, drain, and manual isolation valves.

ALTERNATE TESTING The identified components will be tested in accordance with IWD-5222, Functional Test Requirements, in conjunction with the associated auxiliary feedwater pump at normal operating pressure.

  • S2-002I3 2-33 Rev. 0 12/93
  • RELIEF REQUEST 5 I. IDENTIFICATION OF COMPONENTS Class 1, 2, and 3 pressure retaining bolting II. IMPRACTICAL CODE REQUIREMENTS IWA-5250(a) (2) states, "if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100:"

III. BASIS FOR RELIEF Typically pressure testing required by the Code will identify mechanical connection leakage from gasket or packing sources.

Routinely during the test or following the test, this leakage is evaluated and additional torquing or other corrective measures on the bolting is applied to eliminate the leakage.

Generally though, this type leakage is not active long enough to cause component damage. Additionally some materials are not affected by boric acid wastage. In some instances testing occurs while containment is subatmospheric and just prior to

  • reactor criticality. Requiring that bolting be removed at this time for minor leakage would severely interrupt the normal start up schedule. Additionally the Code requirement does not incorporate previous NRC commitments associated with boric acid wastage on bolting. Surry at the start of a reactor refueling will examine safety-related components inside containment for boric acid leakage. Each component affected by boric acid leakage will be evaluated for potential damage and corrective action at this time. The current Code requirement assumes no examination or evaluation like this has taken place, and defaults conservatively that leakage found has been in place for a significant time period to cause damage. This approach should be considered impractical, when compared to the alternative arrangements proposed .
    • S2-002I3 2-34 Rev. 0 12/93

4lt Relief Request 5 (CON'T)

IV. ALTERNATIVE REQUIREMENTS Bolting in situations requiring removal and visual (VT-3) examination may be limited to one bolt nearest the leakage source. If that bolt has evidence of degradation, then all other bolting in the connection shall be removed and visually (VT-3) examined and evaluated to the Code requirements. The limitation of selecting only one bolt initially is the same as the Code requirements found in the 1992 Edition of ASME Section XI, IWA-5250(a) (2). Bolting removal, however would be limited to bolting material affected by boric acid wastage, such that a visual (VT-3) examination would identify this condition.

Additionally for bolting subject to boric acid wastage examined during pressure tests conducted just prior to start up (Class 1 system leakage, Class 1 hydrostatic, Class 2 pressure tests scheduled in association with the Class 1 tests, and Code Case N-498 tests) in subatmospheric conditions, leakage identified near bolted connections shall be evaluated for removal need based upon the extent of leakage, correction requirements, and previous examination

  • history associated with commitment or any other examinations conducted during that refueling outage. This evaluation shall be subject to the review of the Authorized Nuclear Inservice Inspector (ANII) .
  • S2-002I3 2-35 Rev. 0 12/93

RELIEF REQUEST 6 I. IDENTIFICATION OF COMPONENTS System: Reactor Coolant (RC)

Charging (CH)

Safety Injection (SI)

Components: Pressure retaining bolting located inside containment, which are normally tested in subatmospheric conditions.

ISI Class: 1 and 2 II. IMPRACTICAL CODE REQUIREMENTS Subparagraph IWA-5242(a) requires, for systems borated for the purposes of controlling reactivity, insulation shall be removed from pressure retaining bolted connections for visual examination VT-2.

III. BASIS FOR RELIEF In some cases the reactor coolant system and systems

  • connected to the reactor coolant system are tested in subatmospheric conditions, when the reactor coolant temperature is greater than 500° Fahrenheit. Typically this testing is done just prior to startup. Removing and reinstalling insulation under these conditions is difficult to perform and considered impractical, when compared to the alternate proposal.

IV. ALTERNATIVE REQUIREMENTS It is proposed that bolted connections on Class 1 systems containing boric acid be examined each refueling outage at zero or static pressure. The examination would be performed with insulation removed. Class 2 bolting will be examined similarly once a period. This alternative only applies to systems that are pressure tested under subatmospheric conditions. In addition the Code required testing will still be conducted without removing the insulation. The test will be held at nominal operating pressure for four hours for insulated systems and ten minutes for noninsulated systems prior to performing the visual VT-2 examination .

  • S2-002I3 2-36 Rev. 0 12/93

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT2 INSERVICE INSPECTION PROGRAM FOR COMPONENT SUPPORTS

  • S2-003I3 3-1 Rev. 0 12/93
  • 3
  • 0.

SECTION 3 TABLE OF CONTENTS INSERVICE INSPECTION PROGRAM FOR COMPONENT SUPPORTS 3.1 PROGRAM DESCRIPTION 3.2 PROGRAM EXCLUSIONS

  • S2-003I3 3-2 Rev. 0 12/93
  • 3.0 INSERVICE INSPECTION PROGRAM FOR COMPONENT SUPPORTS 3.1 PROGRAM DESCRIPTION The Inservice Inspection Program for ISI Class 1,2, and 3 component supports meets the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, and the alternative provisions found in Code Case N-491, "Alternative Rules for Examination of Class 1, 2, 3, and MC Component Supports of Light-Water Cooled Power PlantsSection XI, Division 1. 11 3.2 PROGRAM EXCLUSIONS This program does not cover the functional testing of snubbers. Snubbers will be tested to the functional testing requirements described in Technical Specifications.

3.3 COMPONENT SUPPORT RELIEF REQUESTS Component Support relief requests are presented in a numeric sequence with a SH prefix. There are no relief requests at this time .

  • S2-003I3 3-3 Rev. 0 12/93

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2

  • INTERIM RELIEF REQUESTS AND MISCELLANEOUS DOCUMENTATION
  • S2-004I3 4-1 Rev. 0 12/93
  • 4.0 SECTION 4 TABLE OF CONTENTS INTERIM RELIEF REQUESTS AND MISCELLANEOUS DOCUMENTATION 4.1 RELIEF REQUESTS (None at this time) 4.2 Miscellaneous Documentation (None at this time)
  • S2-004I3 4-2 Rev. 0 12/93
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