ML18153A242

From kanterella
Jump to navigation Jump to search

Forwards Safety Evaluation Re Third 10-yr Inservice Insp Program Relief Requests SR-018 Through SR-024
ML18153A242
Person / Time
Site: Surry 
Issue date: 04/07/1998
From: Kuo P
NRC (Affiliation Not Assigned)
To: Ohanlon J
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
Shared Package
ML18153A243 List:
References
TAC-M97686, NUDOCS 9804100254
Download: ML18153A242 (23)
v • d • e


Text

e UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. J. P. O'Hanlon Senior Vice President - Nuclear Virginia !Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard

  • Glen AIIEm, Virginia 23060 April 7, 1998

SUBJECT:

- SURRY.POWER STATION,-UNIT-2 --T-HIRD* 1 Oa.YEAR INSERVICE* INSPECTION-PROGRAM RELIEF REQUEST NOS. SR-018 THROUGH SR-024 (TAC NO. M97686)

Dear Mr. O'Hanlon:

The staff, with technical assistance from the Idaho National Engineering and Environmental Laboratory (INEEL), has completed its evaluation of the information provided by Virginia Electric and Power Company's (VEPCO's) submittal dated January 2, 1997, as supplem,~nted June 2, 1997.

Our evaluation and conclusions are. contained in Enclosure 1, the Safety Evaluation. Based on the information submitted, we.adopt our contractor's conclusions and recommendations presented in Enclosure 2, the Technical Evaluation Letter Report.

The Code requirements for the weld configurations addressed in Relief Request No. SR-018 would represent an undue hardship for the licensee because they would require disassembly of the reactl)r coolant pump flange bolting with excessive radiation exposure to plant personnel.

The subjE:!Ct relief request is granted pursuant to 1 O CFR 50.55a(a)(3)ii as the Code requirements would result in hardship without a compensating increase in the level of quality and safety.

The Code requirements for the weld configurations addressed in Relief Requests Nos. SR-019, SR-020, SR-021, SR-023, and SR-024 are impractical because the configuration precludes access for complete examination, or in the case of SR-023, surface irregularities preclude complete ultrasonic examination. These access and surface problems would require significant modifications to enable a complete examination, which would cause a burden on VEPCO and in one case, make the weld less safe. The subject relief requests are granted, with one condition, 98041002S4 980407 PDR ADOCK 05000281 G

PDR

e pursuant to 10 CFR 50.55a(g)(6)(i) as the Code requirements are impractical and your proposed alternatives provide reasonable assurance of operational readiness for the affected systems; for relief request SR-023, relief is granted provided that the remote visual examinations are performed with color video equipment.

Relief request SR-022 was withdrawn in the June 2, 1997, supplemental submittal.

This concludes our efforts on this issue and we are, therefore, closing out TAC No. M97686.

Docket No. 50-281

Enclosures:

As stated cc w/encls: See next page DISTRIBUTION:

Sincerely, (Original Signed By)

P. T. Kuo, Acting Director Project Directorate 11-1 Division of Reactor Projects - 1/11 DoeketF-ile PUBLIC PDll-1 R/F J. Zwolinski P. T. Kuo E. Dunnington C. Casto G. Edison L. Plisco, Region II T. Harris (e-mail SE only, TLH3)

OGC/ACRS G. Hill (2 copies)

DOCUMENT NAME: G:\\SURRY\\M97686.REL To receive a copy of this document, indicate in the box: "C" = Copy without attachmenUenclosure "E" = Copy.

with attachmenUenclosur "N" = No copy OFFICE PM:PDII-1 LA:PD 11-1 NAME DATE 03/ II /98 03/

/98 Official

~ *----

        • --*--------*****-----~--***-~---

Mr. J. P. O'Hanlon Virginia Electric and Power Company cc:

Mr. Michael W. Maupin, Esq.

Hunton and Williams Riverfront Plaza, East Tower 951 E. Byrd Street Richmond, Virginia 23219 Mr. David Christian, Manager Surry Power Station - * * -- * - -- --- * *

  • Virginia Electric and Power Company 5570 Hog Island Road Surry, Vir~1inia 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. Lc,ugh Virginia State Corporation Commission Division of Energy Regulation _

P.O.Box1197 Richmond, Virginia 23209.

Regional Administrator, Region II U.S. Nuclear Regulatory Commission Atlanta Federal Center 6~ Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303 Robert B. Strnbe, M.D., M.P.H.

State Health Commissioner Office of the Commissioner Virginia Department of Health P.O. Box 2448 Richmond, Virginia 23218 e

Surry Power Station Office of the Attorney General Commonwealth of Virginia 900 East Main Street Richmond, Virginia 23219 Mr. J. H. McCarthy, Manager Nuclear Licensing & Operations

--Support*--**---***---* -- - -----*--

Innsbrook Technical Center Virginia Electric and Power Company 5000 Dominion Blvd.

Glen Allen, Virginia 23060 Mr. R. C. Haag U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. W. R. Matthews, Manager

  • North Anna Power Station P. 0. Box402 Mineral, Virginia 23117

. -****.. --- ____,,,__.....i.......,.~----*'-*-*----------*--*--**-*.. ******* _,.. -----*--*

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN REQUESTS FOR RELIEF

1.0 INTRODUCTION

FOR VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION; UNIT 2 - *.

DOCKET NUMBER: 50-281 In order to demonstrate the operability of ASME Code Class 1, 2, and 3 components, the Technical Specifications (TS) for Surry Power Station, Unit 2, state that the inservice inspection of the Ame!rican Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PV) Ccide and applicable addenda as required by 10 CFR 50.55a(g), except where specific written reliE~f has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authc>rized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the e"tent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservic1~ examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications. listed therein.

The applicable edition of Section XI of the ASME Code for the Surry Power Station, Unit 2 second 10-year inservice inspection (ISi) interval is the 1989 Edition.

9804100258 980407 PDR ADOCK 05000281 G

PDR.

          • - -***----~. -. --** - - *-*-* -~-*-

.. --**--**-... -~---*---**-*~*--*~-- '--------*--**--***-***--..i-.-... ~.-................... _, ___ *---~******

e e Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.

In a letter dated January 2, 1997, as supplemented June 2, 1997, Virginia Electric and Power Company, (licensee), submitted requests for relief (SR-018 through SR-024) to ASME Section XI requirements for Surry Power Station, Unit 2. The licensee withdrew Request for Relief No.

SR-022 in its Jetter dated June 2, 1997.

2.0 EVALUATION The staff, with technical assistance from its contractor, the Idaho National Engineering and Environme,ntal Laboratory (INEEL), has evaluated the information provided by the licensee in support of its third 10-year inservice inspection interval program plan requests for relief for Surry Power Station, Unit 2. Based on the information submitted, the staff adopts the contractor':s conclusions and recommendations presented in the Technical Letter Report (TLR) attached.

Request for Relief SR-018: Section XI, Examination Category 8-J, Item 89.40, Class 1 Socket Welds requires 100% surface examination of the Class 1 socket welds as defined by Figure IWB-2500-B. Pursuantto 10 CFR 50.55a(a)(3)(ii), the licensee proposed to perform the Code:..

required surface examinations on substitute welds in lieu of the Class 1 socket welds listed below.

G't>? :',*:J,Y~lg.,N,o.:'.:i}/;Jit;f;~,, **. :\\}Ji11iu~~ N'§::,:]!(t?,.r :tfi'fl:i1i'}\\*.t?/~f)Jfrg~:~~. >.,*,1\\< *:;i 1-01 2"-CH-397-1502 11548-WMKS-RC-1225 1-01 2"-CH-395-1502 11548-WMKS-RC-1123 1-01 2"-CH-393-1502 11548-WMKS-RC".'1024 The licensee stated:

"Jt is proposed that the substitution with another weld be counted as meeting the Code requirements. In addition:

e e 1. A visual (VT-2) examination will be performed during the normally scheduled system leakage test each refueling outage.

2*. Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3. The containment atmosphere particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4. -If a reactor coolant-pump main flange isdisassembled,-thereby*making the weld -

accessible for examination, the welds will be examined consistent with Code requirements.

The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised."

The Code requires 100% surface examination of the subject seal injection pipe welds.

However, these welds are inaccessible without disassembly of the reactor coolant pump main flange boltlng, which is a major effort that requires many manhours from skilled maintenance personnel. Disassembly of the pump would result in excessive radiation exposure to plant personnel. Therefore, imposition of the Code requirements would cause a burden on the licensee.

The licensee has performed the Code-required surface examination on one of the injection seal pipe welds. In addition, the licensee will select other welds for examination to maintain the 25%

examination sample for Examination Category 8-J welds. Furthermore, the seal injection pipe welds receive the Code-required VT-2 visual examination for leakage each refueling outage.

The combination of these examinations is sufficient for detecting existing patterns of inservice degradation and provides reasonable assurance of the continued structural integrity of the subject welds. Therefore, it is concluded that compliance with the Code requirements would result in hardship without a compensating increase in the level of quality and safety, and the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

Request for Relief SR-019: Examination Category 8-J, Item 89.32, Class 1 Branch Connection Weld requires 100% surface examination, as defined by Figure IWB-2500-5, for branch conn,ection pipe welds less than 4-inch nominal pipe size. The licensee, pursuant to 10 CFR 50.55a(g)(5)(iii), has requested relief from performing the Code-required surface examination on branch connection Weld 4-11 BC on Line 3"-RC-335-1502.

The licensee proposed as an alternative that a surface examination performed on the fillet weld of the reinforcement pad be substituted for the Code surface examination. In addition:

1. "A visual (VT-2) examination will be performed during the normally scheduled system leakage t,;st each refueling outage."
  • 2. "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />."
3. "The containment atmosphere particulate radioactivity i$ checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

"The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised."

  • The-Code! requires*a*100% surface examination of the*subject branch con*nection weld. -*

However, access to this weld is completely restricted by a reinforcing saddle plate which is fillet welded over the pressure-retaining branch connection weld. Therefore, the design with the reinforcin~J saddle plate makes the Code requirement impractical for this branch connection weld. To gain access for examination, the saddle plate would have to be removed and the branch connection redesigned and modified. Imposition of this requirement would cause a burden on the licensee.

In lieu of the Code-required surface examination, the licensee will perform a surface examination of the fillet weld used to attach the saddle plate to the main pipe and branch pipe.

Examination of these welds will detect any gross structural deformation and confirm the overall integrity of the branch connection. In addition, the licensee will perform VT-2 visual examinations of these areas in conjunction with the Class 2 pressure tests. The staff concludes that the alt1~rnative surface examination, along with the Code-required pressure tests, will detect any significant patterns of degradation occurring at the branch connection and will ensure the structural integrity of the subject branch connection. Therefore, Request for Relief No. SR-019 is granted pursuant to 10 CFR 50.55a(g)(6)(i).

Request for Relief SR-020: Section XI, Examination Category 8-F, Item 85.70, Steam Generator Nozzle-to-Safe End Welds requires 100% volumetric and surface examinations, as defined by Figure IWB-2500-8, for steam generator nozzle-to-safe end welds 4-inch nominal pipe size or larger. Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the volumetric coverage requirements of the Code for the steam generator nozzle-to-safe end welds listed in the table below.

1-05DM

/29"-RC-301-2501 R 1"'.06DN _______ _

/31 "-RC-308-2501 R 0

  • 28 55 55 100 Nozzle geometry and surface condition. Machined surface on pipe-side. As-cast surface on nozzle side.

_ 28_ 0 _ 55 __ 55 ____ 100 __ Nozzle geometry and surface condition. Machined surface on pipe-side. As-cast surface on nozzle side.

2-axial scan, 180° from isometric flow direction 5-axial scan, same direction as isometric flow

?-circumferential scan, clockwise 8-circumferential scan, counter-clockwise The licensee proposed as an alternative that the examination already completed at the reduced coverage bH counted as meeting the Code requirements. In addition:

1. "A visual (VT-2) examination will be performed during the normally scheduled system leakage test each refueling outage."
2. "Technical_ Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />."
3. "The containment atmosphere particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

"The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised."

The Code requires 100% volumetric and surface examination of the subject steam generator nozzle-to-safe end welds. However, nozzle geometry and surface conditions make complete volumetric examination impractical for these welds. To meet the Code coverage requirements, the nozzle safo ends *and associated piping would require design modifications to allow access for examination. Imposition of this requirement would create a burden on the licensee.

Approximately 35% (cumulative) of the Code-required volumetric examination and 100% of the Code-required surface examination was obtained for the tv,o steam generator* safe end welds:

In addition, there are other safe-end welds in the reactor coolant system that are receiving complete volumetric examination. The combination of the complete surface examination, partial volumetric examination, and the complete examination of other similar welds assures

    • -- *l*,**-* -*** *
  • -*----- --~-

<,A*--------~---~--***.-~-----

  • that existing patterns of degradation will be detected. As a result, reasonable assurance of the structural integrity of the dissimilar metal safe end welds has been provided. Therefore, Request for Relief No. SR-020 is granted pursuant to 10 CFR 50.55a(g)(6)(i).

Request for Reiief SR-021: Section XI, Examination Category 8-J, Item 89.11, Class 1 Circumferential Weld 2-30 requires 100% volumetric and surface examination, as defined by Figure IWB-2500-5, for Class 1 circumferential welds 4-inch nominal pipe size and larger. The licensee pursuant to 10 CFR 50.55a(g)(5)(iii), requested relief from performing the surface examination, to the extent required by the Code, for Weld 2-30 on line 4"-RC-315-1502. Details are listed in Table SR-021-1 below.

2-30 81%

Whip restraint on piping extending across the weld at 0°, 90°, 180° and 270°.

The licensee proposed as an alternative that the examination already completed at the reduced coverage be counted as meeting the Code requirements. In addition:

1. "A visual (Vf-2) examination will be performed during the normally scheduled system leakage test each refueling outage."
t "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at lea~t once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />."
3. "The containment atmosphere. particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

The Code requires 100% surface and volumetric examination of the subject circumferential pipe weld. Howev,:ir, access to the weld is restricted by pipe whip restraints that extend across the weld and prevent completion of the Code-required surface examination. Therefore, the Code coverage requirements are impractical for this weld. To meet the Code requirements, the whip restraint and portions of the piping system would have to be redesigned and modified.

Imposition of this requirement would result in a burden on the licensee.

The licensee completed 81 % of the Code-required surface examination and essentially 100% of

. the volumetric examination. In addition, this weld is part of a larger sample of Class 1 circumferential welds that received complete examination. The combination of the volumetric e:;,:amination, the surface examination to the extent practical, and the examination of other Class 1 piping welds i3ssures that existing patterns of degradation will be dete:ted. As a result, reasonable assurance of the structural integrity of the subject weld has been provided.

Therefore, Request for Relief No. SR-021 is granted pursuant to 10 CFR 50.55a(g)(6)(i).

'"7 -

Request: for Relief SR-022: Section XI, Examination Category C-G, Item C6.10, Safety Injection Pump Casing Weld. In the licensee's June 2, 1997, response to the NRC's RAI, the licensee withdrew Request for Relief SR-022.

Request for Relief SR-023: Section XI, Examination Category B-D, Item 82.120, Pressurizer Nozzle Inside Radius (IR) Sections requires 100% volumetric examination, as defined by Figure IWB-2500-7, of pressurizer nozzle inside radius sections. Pursuant to 10 CFR 50.SSa(g)(S)(iii),

the licensee requested relieffrom examining pressurizer nozzle IR sections 1 ONIR and 11 NIR to the extl3nt required by the Code.

The Code requires* 100%-volumetric**examination of the subject nozzle inside* radius* (IR) sections. However, the nozzles are integrally cast and the s*urface conditions preclude complete ultrasonic examination. Reducing the surface irregularities by grinding would reduce the wall thickness, which is not technically prudent. Therefore, the Code-required volumetric examination is impractical for the pressurizer iriside radius sections at Surry, Unit 2. To provide.

a surface suitable for ultrasonic examination, the pressurizer nozzles*would require design

  • modification. Imposition of this requirement would result in a considerable burden on the licensee.

In addition to performing the Code-required volumetric examination to the extent practical, the licensee proposed to perform a VT-1 visual examination of the inside surface of the IR sections.

This alternc1tive is capable of detecting any significant patterns of degradation and will provide reasonaple assurance of the structural integrity of the IR sections provided that remote visual examinations are *performed with color yideo equipment. Therefore, Request for Relief No. SR.;

023 is granted pursuant to 10 CFR 50.55a(g)(6)(i), provided that remote visual examinations are performed with color video equipment.

  • Request for Relief SR-024: Examination Category 8-J, Item 89.11, Class 1 Circumferential Piping Weld 1-03 requires 100% volumetric and surface examination, as defined by Figure IWB-2500-5, for Class 1 circumferential welds 4-inch nominal pipe size and larger. The licensee requested C1ode relief pursuant to 10 CFR 50.SSa(g)(S)(iii), from performing volumetric examination, to the extent required by the Code, for Weld 1-03 on lirie 29"-RC-307-2501 R.

Details are listed in Table SR-024-1 below.

  • 1-03 0

100 25 25 100 Pipe/valve configuration prevents scanning from side 2. A 60° angle was used to increase coverage (direction 5).

Because of use of longitudinal waves, only Ya node examinations were possible.

2.:.axial scan, 180° from isometric flow direction 5-axial scan, same direction as isometric flow I ?-circumferential scan, clockwise 1 B-circumferential scan, counter-clockwise The licensee proposed as an alternative (as stated):

"It is proposed that the examination already completed at the reduced coverage be counted as meeting the Code requirements. In addition:

1.

A visual (VT-2) examination will be performed during the normally scheduled system leakage test each refueling outage.

2.

Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.

The containment atmosphere particulate radioactivity is checked every

  • 12 hours.

The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised."

The Code requires 100% surface and volumetric examination of the subject circumferential pipe weld. Howev1er, the weld configuration restricts coverage and precludes 100% volumetric examination. Therefore, the Code coverage requirements are impractical for this weld. To meet the Codt; requirements, the weld joint would have to be redesigned and modified.

Imposition of this requirement would result in a burden on the licensee.

The licensee examined approximately 37% of the Code-required volume and 100% of the surface. In addition, this weld is part of a larger sample of Class 1 circumferential welds that did receive compleite examination. The combination of the volumetric examination to the extent practical, the ciomplete surface examination, and the examination of other Class 1 piping welds assures that existing patter.-is of degradation will be detected. As a result, reasonable assurance of the structural integrity of the subject weld has been provided. Therefore, Request for Relief No. SR-024 is granted pursuant to 10 CFR 50.55a(g)(6)(i).

e 3. CONCLUSIONS The staff has reviewed the licensee's submittal and concludes that for Requests for Relief SR-019, 020, 021, and 024, the requirements are impractical at Surry Power Station, Unit 2 and that reasonable assurance of the structural integrity is provided by the examinations performed.

Therefore, relief is granted pursuant to 1 O CFR 50.55a(g)(6)(i) for these requests. For Request for Relief SR-023, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i), provided the remote visual examinations are performed with color video equipment.

For Request for Relief SR-018, imposition of the Code requirements would cause a burden without a-.compensating increase in qua1ity *and safety. -- Therefore; the licens-ee's t>roposed_ --

  • alternativei is authorized pursuant to 10 CFR 50.55a(a)(3){ii).

Request ft>r Relief SR-022 was withdrawn by the licensee in their June 2, 1997, supplemental submittal.

TECHNICAL LETTER REPORT THIRD 10-YEAR INTERVAL INSERVICE INSPECTION

1.0 INTRODUCTION

REQUESTS FOR RELIEF VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT 2 DOCKET NUMBER 50-281 By letter dated January 2, 1997, the licensee, Virginia Electric Power Company,

- submitted *seven-requests for relief from the -requirements of Section XI of thef Americari Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. In response to a Nuclear Regulatory Commission (NRC) request for additional information (RAI), the licensee provided further information in a letter dated June 2, 1997. The licensee also withdrew Request for Relief SR-022 in that letter. The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the information provided by the licensee in support of these requests for relief in the following section.

2.0 EVALUATION The Code of record for the Surry Power Station, Unit 2, third 10-year inservice insp,ection (ISi) interval, which began May 10, 1994, is the 1989 Edition of ASME Code,Section XI.

A)

Request for Relief SR-018, Examination Category s.:.J, Item 89.40, Class 1 Socket Welds

!8ode Requirement: Examination Category 8-J, Item 89.40 requires 100% surface 1:!xamination of the Class 1 socket welds as defined by Figure IWB-2500-8.

l.icensee's Proposed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed to perform the Code-required surface examinations on substitute welds in lieu of the Class 1 socket welds listed below.

[i*>,) ;;~~~1~;t:.Jp:*>::~:fi):f,J:{i_\\iK~,'.,;'.µir1~.:~e>;;_t:_...,

f * <..-*-::;1:_;'°<< i;>rawing *f\\Jo*~. :1:~: :JJ 1-01 2"-CH-397-1502 11548-WMKS-RC-1225 1-01 2"-CH-395-1502 11548-WMKS-RC-1123 1-01 2"-CH-393-1502 11548-WM KS-RC-1024 The licensee stated:

"It is proposed that the substitution with Jnother weld be counted as meeting the Code requirements. In addition:

- 1. "A visual (VT-2) examination will be performed during the normally scheduled system leakage test each refueling outage.

2. "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3. "The containment atmosphere particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

-4, -"-If a reactor-coolant-pump main -flange-is disassembled; thereby making the weld accessible for examination, the welds will be examined consistent with Code requirements.

"The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised."

Licensee's Basis for the Proposed Alternative (as stated):

"The components listed above are terminal ends on seal injection piping connected to the reactor coolant pumps. The welds are completely inaccessible due to their closeness to the pump flange (i.e., they are underneath the flange. The attached picture shows the configuration.) However, when the reactor coolant pump main flange is disassembled, the weld is accessible for examination. However, disassembly of the pump would cause a burden without a compensating increase in quality and safety. All terminal end welds are scheduled for examination.

Therefore, substitution with another weld is not feasible. However, another weld will be selected for examination in order to ensure that 25% of examination category 8-J will be examined."

In the June 2, 1997, letter, the licensee stated:

"During the last refueling outage (March 1997), the "A" reactor coolant pump (RCP) main flange bolting was disassembled for maintenance making the seal injection pipe weld accessible for examination, and the weld was examined. No indications WBre identified, and the weld was accepted. However, future partial disassembly of reiactor coolant pumps for the sole purpose of examining the subject seal injection pipe welds would represent a considerable hardship. Shielding would be required to be installed in the motor room to reduce radiation levels. In addition, it would take two individuals at least twenty-four hours to remove and reinstall enough RCP bolting to make the pipe weld accessible. Three to four bolts would also have to be removed in the location of the piping weld, as well as another three or four bolts 180° across from the bolts in the weld area. Furthermore, loosening the bolting could cause gasket problems which would require removal of the remaining bolts, repracement of the gasket, and retensioning of the bolting. Past RCP maintenance activities at Surry indicate that it can take up to six days to properly retention bolting on the RCPs. Reasonable assurance of system structural integrity is provided by

~-- *---------~ *-'-**---****-- *-***-- - *~---*--*~---*--J.. __ ~------...---------. -

. _________....._.....__. ___.~--~------*--~---

e the addition of another weld to the 8-J welds examinations to makeup the 25%

sample, the system leakage test performed every refueling outage, primary system leakage monitoring during routine operation, and the containment atmospheric radioactivity monitoring system. These assurances are included in Relief Request, SR-018.

"Consequently, we request relief from performing the weld examination coverage requirements for the subject weld pursuant to 10 CFR 50.55a(a)(3)(ii) for the reasons cited in our January 2, 1997 response as well as the supplemental basis provided above. n

~-

    • -*-*r*-

Evaluation: The Code requires 100% surface examination of the subject seal injection pipe welds. However, these welds are inaccessible without disassembly of the RCP main flange bolting, which is a major effort that requires many manhours from skilled maintenance personnel. Disassembly of the pump would result in excessive radiation exposure to plant personnel. Therefore, imposition of the Code requirements would cause a burden on the licensee.

The licensee has performed the Code-required surface examination on one of the injection seal pipe welds. In addition, the licensee will select other welds for examination to maintain the 25% examination sample for Examination Category B-J.

welds. Furthermore, the seal injection pipe welds receive the Code-required VT,-2 visual examination for leakage each refueling outage. The combination of these examinations is sufficient for detecting existing patterns of inservice degradation and provides reasonable assurance of the continued structural integrity of the subject welds. Therefore, it is concluded that compliance with the Code requirements would result in hardship without a compensating increase in the level of quality and safety, and it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

B)

Reguest for Relief SR-019, Examination Category B-J, Item 89.32, Class 1 Branch Connection Weld Cc>de Reguirement: Examination Category 8-J, Item 89.32 requires 100% surface examination, as defined by Figure IWB-2500-5, for branch connection pipe welds less than 4-inch nominal pipe size.

Licensee's Code Relief Reguest: Pursuant to 10 CFR 50.SSa(g)(S)(iii), relief is requested from performing the Code-required surface examination on branch connection,Weld 4-11 BC on Line 3"-RC-335-1502.

Licensee's Basis for Reguesting Relief (as stated):

"The weld listed above is covered by a reinforcement pad/saddle weld, which totally covers the examination area described in the Code. Therefore, examination coverage cannot be achieved due to the configuration of the piping. This weld is a

  • J --*
  • -~-_,..___.. ______ **-**-~-. *--~-----------
  • -----------M"-*-*--------* ** * **

e e branch connection and, as such, all welds of this type are scheduled for examination. Therefore, substitution with another weld is not feasible."

Licensee's Proposed Alternative (as stated):

"It is proposed that a surface examination performed on the fillet weld of the reinforcement pad be substituted for the Code surface examination. In addition:

  • 1. "A visual (VT-2) examination will be performed during the normally scheduled

_ __syst_ern _l~a~a_g~ t~~t _eac~ _refueling o_uta_ge, _

2. "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3. "The containment atmosphere particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised."

.Evaluation: The Code requires a 100% surface examination of the subject branch

,connection weld. However, access to this weld is completely restricted by a reinforcing saddle plate which is fillet welded over the pressure-retaining branch connection weld. Therefore, the design with the reinforcing saddle plate makes the Code requirement impractical for this branch connection weld. To gain access for examination, the saddle plate would have to be removed and the branch connection redesigned and modified. Imposition of this requirement would cause a burden on the licensee.

In lieu of the Code-required surface examination, the licensee will perform a surface examination of the fillet weld used to attach the saddle plate to the main pipe and

~ranch pipe. Examination of these welds will detect any gross structural dt~formation and confirm the overall integrity of the branch connection. In addition, the licensee will perform VT-2 visual examinations of these areas in conjunction with the Class 2 pressure. tests. The INEEL staff concludes that the alternative surface examination, along with the Code-required pressure tests, will detect any si~Jnificant patterns of degradation occurring at the branch connection and will ensure the structural integrity of the subject branch connection. Therefore, it is re~;ommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

  • -----*-*- -*--- *- _,. ___. ____,....__~,_... __.,,._...._.._..._.....

___ _.._.._ _____ ~---

e C)

Request for Relief SR-020, Examination Category B-F, Item B5.70, Steam Generator Nozzle-to-Safe End Welds Code Requirement: Examination Category B-F, Item B5.70 requires 100%

volumetric and surface examinations, as defined by Figure IWB-2500-8, for steam generator nozzle-to-safe end welds 4-inch nominal pipe size or larger.

Licensee's Code Relief Request: Pursuant to 1 O CFR 50.55a(g)(5)(iii), relief is requested from the volumetric coverage requirements of the Code for the steam generator nozzle-to-safe end welds listed in the table.below:

1-05DM 0

28 55 55 100

/29"-RC-301-2501 R 1l-06DN 28 0

55 55 100

/31 "-RC-308-2501 R 2-axial scan, 180° from isometric flow direction 5--axial scan, same direction as isometric flow 7.. circumferential scan, clockwise 8-circumferential scan, counter-clockwise Nozzle geometry and surface condition. Machined surface on pipe-side. As-cast surface on nozzle side.

Nozzle geometry and surface condition. Machined surface on pipe-side. As-cast surface on nozzle side.

Licensee's Basis for Requesting Relief (as stated):

"The components listed above have been examined to the extent practical as required by the Code. However, full volumetric coverage could not be achieved due to joint configuration and material characteristics. Coverage of the volumetric and surface examinations is detailed in Tables SR-020-1 and SR-020-2

[paraphrased above]. Figure SR-020-1* is provided as graphic detail of the limitations experienced. Substitution with another weld is not feasible because all welds in the Category and Item must be examined.

Licensee's Proposed Alternative (as stated):

"It is proposed that the examination already completed at the reduced coverage be counted as meeting the Code requirements. In addition:

  • Contained in licensee's submittal but not this report.

_..,. __. ** -- --- _.., ~--.......... - *~.,.- ***---..-..v-.-...--_.._.._ _______,, _____._._ ____._~~----"-*-... ~~---*-..

e 1. "A visual (VT-2) examination will be performed di.iring the normally scheduled system leakage test each refueling outage.

2. "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3. "The containment atmosphere particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
  • * - *- - * *"The proposed alternative*examinations*stated.above will ensure*that the overall***

level of plant quality and safety will not be compromised."

Evaluation: The Code requires 100% volumetric and surface examination of the subject steam generator nozzle-to-safe end welds. However, nozzle geometry and surface conditions make complete volumetric examination impractical for these welds. To meet the Code coverage requirements, the nozzle safe ends and associated piping would require design modifications to allow access for

,examination. Imposition of this requirement would create a burden on the licensee.

Approximately 35% (cumulative) of the Code-required volumetric examination and

'I 00% of the Code-required surface examination was obtained for the two steam generator safe end welds. In addition, there are other safe-end welds in the reactor coolant system that are receiving complete volumetric examination. The combination of the complete surface examination, partial volumetric examination, and the complete examination of other similar welds assures that existing patterns of degradation will be detected. As a result, reasonable assurance of the structural integrity of the dissimilar metal safe :end welds has been provided. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

D)

Reguest for Relief SR-021. Examination Category 8-J, Item 89.11, Class 1 Circumferential Weld 2-30 Code Reguirement: Examination Category 8-J, Item 89.11 requires 100%

volumetric and surface examination, as defined by Figure IW8-2500-5, for Class 1 circumferential welds 4-inch nominal pipe size and larger.

Licensee's Code Relief Reguest: Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the surface examination, to the extent required by the Code, for Weld 2-30 on line 4"-RC-315-1502. Details are listed in Table SR-021-1 below.

...,. ---* ----*--**--'-~*---'--*-~---*------*..£...*--...-.-~---------- ~~--~---,--. -*~...... --~--'"---*---*...

e e* *:,~,;~i~~¥~~g,~~sRgo2~~~1r.

, -s~\\~~[ Iii~~~f ~f @~1l? f ~~~lttta~--~}.,

2-30 93.3%

81%

Whip restraint on piping extending across the weld at 0°, 90°, 180° and 270°.

Licensee's Basis for Reguesting Relief (as stated):

"The component listed above has been examined to the extent practical as required

-- --.. by the Code. However, fun surface coverage-could not be-achieved due to --

interferences from an adjacent pipe support. The required volumetric examination had limitations, but they were less than 10%. Coverage of the volumetric and surface examinations is detailed in Table SR-021-1. Picture SR-021-1* is provided as graphic detail of the limitations experienced. Substitution with another weld of the same size would not necessarily improve the examination coverage since this weld is representative of plant design and similar geometric conditions are

,expected.

Licensee's Proposed Alternative (as stated):

"!tis proposed that the examination already completed at the reduced coverage be counted as meeting the Code requirements. In addition:

1. "A visual (VT-2) examination will be performed during the normally scheduled system leakage test each refueling outage.
2. "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3. "The containment atmosphere particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

Evailuation: The Code requires 100% surface and volumetric examination of the subject circumferential pipe weld. However, access to the weld is restricted by pipe whip restraints that extend across the weld and prevent completion of the Code-required surface examination. Therefore, the Code coverage requirements are impr'actical for this weld. To meet the Code requirements, the whip restraint and portions of the piping system would have to be redesigned and modified.

Imposition of this requirement would result in a burden on the licensee.

The licensee completed 81 % of the Code-required surface examination and essentially 100% of the volumetric examination. In addition, this weld is part of a

  • Contained in licensee's submittal but not this report.

*~** - -~---4----~ __,, _

_____. ____,. __... ___ =-~-... -.......L.---..--~-*

, -~-.... ~-*-.... -~-------**-~------.

.,. 8-larger sample of Class 1 circumferential welds that received complete examination.

The combination of the volumetric examination, the surface examination to the extent practical, and the examination of other Class 1 piping welds assures that existing patterns of degradation will be detected. As a result, reasonable assurance of the structural integrity of the subject weld has been provided. Therefore, it is recommended that relief be granted pursuant to 1 O CFR 50.SSa(g)(S)(i).

E.

Reguest for Relief SR-022, Examination Category C-G, Item CS.10, Safety Injection Pump Casing Weld

-Note:-ln -the June 2-, 1997-, -response to the NRC1s-RAI, the -licensee withdrew - - * * -

Request for Relief SR-022.

F.

Reguest for Relief SR-023, Examination Category B-D, Item 82.120, Pressurizer Nozzle Inside Radius (IR) Sections Code Reguirement: Examination Category 8-D, Item 82.120 requires 100%

volumetric examination, as defined by Figure IWB-2500-7, of pressurizer nozzle inside radius.sections..

!Jcensee's Code Relief Reguest: Pursuant to 1 O CFR 50.55a(g)(5)(iii), the licensee requested relief from examining pressurizer nozzle IR sections 1 ONIR and 11 NIR to the extent required by the Code.

  • Licensee's Basis for Reguesting Relief (as stated}: -

"The components listed above have been examined to the extent practical as required by the Code. However, full volumetric coverage could not be achieved due to inherent interferences from a rough and bumpy surface, which causes the ultrasonic search unit to lose contact and change angles. The use of alternate angles would not have improved volumetric coverage. The nozzles are integrally cast to the pressurizer, therefore weld contour preparation (grinding or weld buildup and grinding) prior to examination will result in vessel wall reduction and is not rec:ommended. Picture SR-023-2 (NIR-11 r is provided as graphic detail of the pressurizer nozzles, and is representative of existing conditions. Substitution with another nozzle inside radius section is not feasible because all of the pressurizer nozzle inside radius sections must be examined."

In the June 2, 1997, response to the NRC's request for additional information, the licensee stated:

  • "It is estimated that only five percent of the required volume could be examined for the r:eason stated in the relief request. The visual examination described in paragraph V of the relief request will be a VT-1 examination. The use of remote
  • Contained in licensee's submittal but not this report.

e e visual aids as allowed by IWA-2211 (c) of ASME Section Xl-1989 Edition may be required."

Licensee's Proposed Alternative (as stated):

"It is proposed that the examination already completed at the reduced coverage be counted as meeting the Code requirements. In addition,

1. "The nozzle inside radius sections will be visually examined from the inside using direct or remote techniques prior to the end of the inspection interval.
2. "A visual (VT-2) examination will be performed during the normally scheduled system leakage test each refueling outage.
3. "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4. "The containment atmosphere particulate radioactivity is checked every

. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised."

Evaluation: The Code requires 100% volumetric examination of the subject nozzle inside radius (IR) sections. However,.the nozzles are integrally cast and the surface conditions preclude complete ultrasonic examination. Reducing the surface irregularities by grinding would reduce the wall thickness, which is not technically prudent. Therefore, the Code-required volumetric examination is impractical for the pressurizer inside radius sections at Surry, Unit 2. To provide a surface suitable for ultrasonic examination, the pressurizer nozzles would require design modification.

lmpo.sition of this requirement would result in a considerable burden on the licensee.

In addition to performing the Code-required volumetric examination to the extent practical, the licensee proposed to perform a VT-1 visual examiriation of the inside surface of the IR sections. This alternative is capable of detecting any significant patterns of degradation and will provide reasonable assurance of the structural int~grity of the IR sections provided that remote visual examinations are performed with color video equipment. Therefore, it is recommended that relief be granted pur:suant to 10 CFR 50.55a(g)(6)(i) provided that remote visual examinations are performed with color video equipment.

~* **-*- --- *------~----

_% _____ -~~-------*...._. ____ ----,=-=-~------------------------*---------------------~~----=--~~---------* _-_________ -___ -___ -___ -___ -_.-__ -__ -__ -_, __ -__ -___ -_-_-___ -_-_. --------,

e e G.

Reguest for Relief SR-024, Examination Category 8-J, Item 89, 11, Class 1 Circumferential Piping Weld 1 ~03 Code Reguirement: Examination Category 8-J, Item 89.11 requires 100%

volumetric and surface examination. as defined by Figure IWB-2500-5, for Class 1 circumferential welds 4-inch nominal pipe size and larger.

Licensee's Code Relief Reguest: Pursuant to 1 O CFR 50.55a(g)(p)(iii), the licensee requested relief from performing the volumetric examination. to the extent required by the Code, for Weld 1-03 on line 29"-RC-307-2501R. Details are listed in Table

-SR-024-1 below.----

1-03 0

100 25 25 100 Pipe/valve configuration prevents scanning from side 2. A 60° angle was used to increase coverage (direction 5).

Because of use of longitudinal waves, only Yz node examinations were possible.

2--axial scan, 180° from isometric flow direction 5-axial scan, same direction as isometric flow 7-circumferential scan, clockwise 8-,::ircumferential scan, counter-clockwise Licensee's Basis for Reguesting Relief (as stated):

"The component listed above has been examined to the extent practical as required

_ by the Code. However, full volumetric coverage could not be' achieved due to joint configuration. Coverage of the volumetric and surface examinations is detailed in Table SR-024-1 [paraphrased above]. Figure SR-024-1" is provided as graphic detail of the limitations experienced. Substitution with another weld of the same size wpuld not necessarily improve the examination coverage since this weld is representative of plant design and similar geometric conditions are expected."

Licensee's Proposed Alternative (as stated):

  • Contained in licensee's submittal but not this report.

t, e

  • "It is proposed that the examination already completed at the reduced coverage be counted as meeting the Code requirements. In addition:
  • 1. "A visual (VT-2) examination will be performed during the normally scheduled*

system leakage test each refueling outage.

2. "Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3. "The containment atmosphere particulate radioactivity is checked every 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:** **- * -- * -- ** *** * ***...

"The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be-compromised."

Evaluation: The Code requires 100% surface and volumetric examination of the subject circumferential pipe weld. However, the weld configuration restricts cc,verage and precludes 100% volumetric examination. Therefore, the Code cc,verage requirements are impractical for this weld. To meet the Code requirements, the weld joint would have to be redesigned and modified. Imposition of this requirement would result in a burden on the licensee.

The licensee examined approximately 37% of the Code-required volume and 100%

of the surface. In addition, this weld is part of a larger sample of Class 1 circumferential welds that did receive complete examination. The combination of the volumetric examination to the extent practical, the complete surface examination, and the examination of other Class 1 piping welds assures that existing patterns of degradation will be detected. As a result, reasonable assurance of the structural integrity of the subject weld has been provided.. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

3.0 CONCLUSION

The INEEL staff has reviewed the licensee's submittal and concludes that for Requests for Relief SR-019, 020, 021, and 024, the requirements are impractical at Surry Power Station, Unit 2 and that reasonable assurance of the structural integrity is provided by the examinations performed. The!refore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i) for these requests. For Request for Relief SR-023, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i) provided remote visual examinations are performed with color video equipment.

For Request for Relief SR-018, imposition of the Code requirements would cause a burden without a compensating !ncrease in quality and safety. Therefore, it is recommended that the

. licensee's propo:sed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii):

Request for Relief SR-022 was withdrawn by the licensee in their June 2, 1997, response to the NRC RAI.

Retrieved from "https://kanterella.com/w/index.php?title=ML18153A242&oldid=4986651"