ML20070R286

Forwards Inservice Insp Relief Requests for Second 10-yr Insp.Nrc Acceptance of ASME Section XI Relief Requests in Attachments 2 & 3 to Enable Effective Implementation of Inservice Insp Plan Requested

ML20070R286
Person / Time
Site:	Point Beach
Issue date:	01/19/1983
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Clark R, Harold Denton Office of Nuclear Reactor Regulation
References
TAC-48762, NUDOCS 8301270395
Download: ML20070R286 (38)

Text

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l O " POWER COMPANY 231 W. MICHIGAN, P.O. E ;X 2046, MILWAUL.l, WI 53201 January 19, 1983 Mr.-H. R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Mr. R. A. Clark, Chief Operating Reactors Branch 3 Gentlemen:

DOCKET NOS. 50-266 AND 50-301 INSERVICE INSPECTION PLAN POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 As required by 10 CFR 50.55a(g) (5) (iii) , Wisconsin Electric Power Company (Licensee) is providing herewith the inservice inspection relief requests for the second ten-year inspection for Point Beach Nuclear Plant, Unit 2. In a letter dated October 6, 1981, Licensee notified the NRC of its plans to develop the Unit 2 long-term inservice inspection plan including the applicable code editions and addenda to be used for this tipdate. In Mr. Colburn's letter dated October 21, 1981, the NRC staff indicated that these proposed plans were acceptable.

The Unit 2 inservice inspection plan has subsequently been developed i

and approved by the Licensee. As is-the case with the Unit 1 plan, l the detailed Unit 2 long-term plan for the second ten-year inservice

, inspection interval is'available for review at the Point Beach facility.

! As you know, due to the vintage of Point Beach Nuclear l . Plant, Unit 2 (the unit went into commercial service on October 1, 1972), the piping systems were designed prior to the concepts of l Code Class 1, 2, and 3 and Safety Class 1, 2, and 3 systems now l

.being used industry wide for newer plants. The pressure vessels at Point Beach Nuclear Plant, which are within the scope of the Unit 2 inservice inspection plan, were~ designed and constructed to Class A, B, or C standards of ASME Section III. The piping systems we"e designed and constructed in accordance with USAS B31.1.

The designY criteria utilized for Point Beach Nuclear Plant, Unit 2 wer$' derived from the then proposed general design criteria of 10 CFR 'iO as clarified in the Point Beach Nuclear Plant Final Safety Ana' lysis Report.

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Mr. H. R. Denton - January 19, 1983

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4 As stated in. Paragraphs 5.2.1 and 5.2.2 of Attachment 1, 4 , - Regulatcry. Guide . l.26 and ANSI Standard - N18.2_ wer: _ utilized as guidance'to determine Code Class 1,:2, and 3 boundaries;for

-inservice.inspectidn purposes. .Since Point Beach _ Nuclear Plant, Unit 2 :was not originally designed to tlut specific requirements of-Regulatory. Guide 1.26.and ANSI Standard N18.2, the. Code Class boundaries established for inservice inspectica purposes meet the

. intent.of theJcriteria established in these standards. The Point

. Beach Nuclear Plant Final Safety Analysis Report was the basis upon which' determination of the-Code Class boundaries

was performed since this documentiis the foundation for the design and safety analysis of tho' unit. Attachment.1 provides a description of the, Unit 2 inservica inspection second ten-year' interval plan.

t Unit 2 relie'f re'questsifor specific ASME Section XI ,

I code requirements are:provided"in Attachment 2. As the second..

-inspection interval progresses and other problem areas are identi-fled, additional' relief req'uests-may need to be submitted.

~

Attachment 3 contains relief requests 6, 7, 8, and 9 for Point Beach Unit 1. .These relief requests are an addition to i' _those submitted.in our. August 20,-1982 letter. These additional requests were determined to be necessary_either as-the. result of fully-implementing the-Unit-1 plan during the 1982 refueling.

outage,or were identified during the finalization of the Unit 2

<- long-term plan.

e As discussed via telephone with Mr. Tim Colburn and

' -other members of your staff on September- 14, 1982, the resolution of' inconsistencies ~concerning two of the.10 CFR Section 50.55a reliefs granted by,Mr. Clark's letter of' August 31,~1982~ has not ,

yet been received. This. resolution is a-necessary part of

. Wisconsin Electric's attempt to satisfactorily cloce out the

' Point. Beach Unit'2 first. ten-year inservice inspection interval 1

requirements.- Your prompt resolution of these inconsistencies

' is requested.

t i We hereby request NRC's. acceptance of the ASME Section i XI relief requests presented in_ Attachments 2 and 3 so that the inservice inspection plan can~be effectively implemented. It is the' intention of Wisconsin Electric to perform examinations in

accordance with this-approved plan and it will be the basis for.

- ' compliance with. regulatory requirements.

Very truly yours, gsu'f  :

Vice Presid t- clear Power

[ C. W. Fay-4 Attachments

! ' Copy to NRC . Resident Inspector s

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ATTACHMENT 1 i' POINT BEACH NUCLFJR PLANT

• l UNIT 2 ISI LONG-TERM PLAN (LTP)

DEVELOPMENT BASIS 4

1.0' PURPOSE -

'Ib present a sununary of the Unit .2 ISI LTP.

2.0 DESCRIPTION

OF LTP 2.1 The LTP provides a listing of all areas required to be examined under

. Articles IWB, IWC, IWD, and IWF of ASMS Section XI.

2.2 .The LTP provides a planning schedule of when a given examination will 3

-be performed during the second 10-year interval.

2.3 The LTP will pr. ovide an updated status of all required ISI.

> 3.0 APPLICABLE CODE EDITION / ADDENDA 3.l' All ISI work will-be conducted in accordance with the 1977/Sununer 1979 Edition / Addenda of ASME Section XI with the exception of Class 1

and 2 piping welds.

'3.2 . The extent of examination (i.e. , frequency, number of welds) of Class 1 and 2 pipirig welds will be determined in accordance with .the 1974/ Summer 1975 Edition / Addenda of ASME Section XI.

3. 3' -The examination methods, acceptance standards, and other aspects of

'the e - ination of Class 1 and 2 piping welds will be in accordance with the 1977/ Summer 1979 Edition / Addenda of ASME Section XI.

I- 24.0- BASIS FOR CODE SELECTION,

-4.1 10 CFR 50-4.1.1 In accordance with 10 CFR 50.55a(g) (4) (ii) , inservice examin-L- ation of components and system pressure tests conducted

(' during the second and successive 10-year intervals shall comply with the requirements of the latest edition of the Code incorporated by reference in 10 CFR 50.55a(b) (2) 12 months prior to the start of the 120-month inspection interval.

4.1.2~ In accordance with 10 CFR 50.55a(b) (2) (ii) and (b) (2) (iv) ,

WE chose to determine the extent of examination of Class 1

'and 2 piping welds in accordance with the 1974/ Summer 1975 Edition / Addenda of ASME Section XI. -

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-Page 2 5

i 4.1.3 The if74/ Summer 1975 Ed. tion / Addenda of ASME Section XI must be utilized to. determine the extent of examination of Class 2 piping welds in the residual heat removal and the tsfety injection systems.

4.2 In the' Fay to Denton letter dated 10/06/81, WE informed the NRC that

'the Units 1 and 2 LTP's would be written to the Codes referenced in 4 3.a, 3.b, and 3.c.

4.3 In the Colburn to Burstein letter, dated 10/21/81, the NRC accepted the WE plans stated in the 10/06/81 letter.

5.0 CODE CLASS BOUNDARIES 5.1 For the purposes of 'ISI, determination of Code class boundaries on

-systems and equipment designated as nuclear safety related and impor-

~

tant to safety were established using the following selection criteria.

5.2 Selection Criteria References 4

5.2.1 Regulatory Guide 1.26 (Revision 2) , " Quality Group Classifi-cations and Standarde for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants."

5.2.2 ANSI Standard N18.2 (1973) and N18.2a (1975), " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."

5.2.3 PBNP Final Safety Analysis Report

a. Section 4 - Reactor Coolant System

b. Section 6 - Engineered Safety Features

c. Section 9 - Auxiliary and Emergency Systems

d. Section 11 - Waste Disposal and Radiation Protection

e. Section 14 - Safety Analysis

f. Appendix A - Seismic Design for Structures and Equipment 5.3 Exemptions 5.3.1 Class 1

a. Atticle IWB-1220 of the 1977/ Summer 1979 Edition / Addenda of ASME Section XI for all components except piping welds.

b. Article IWB-1220 of the 1974/Sununer 1975 Edition / Addenda of ASME Section XI for piping welds.

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6 Q Attachment 1 Page,3 5.3.2 class 2

a. Article IWC-1220 of the 1977/Sununer 1979 Edition / Addenda of ASME Section'XI for all components except piping,

b. Article IWC-1220 of the'1974/ Summer 1975 Edition / Addenda of ASME Section XI for piping.

c. Due to the nonacceptance by the IGC of exemption criteria IWC-1220(c) of the 1974/Stanmer 1975 Edition / Addenda of ASME Section XI at other plants, this exemption was not

! taken. This paragraph permits exemption of components which perform an emergency core cocling function, provided the control of the chemistry of the contained fluid is verified by periodic sampling and testing.

d. The suction piping for the safety injection, containment spray, and residual heat removal pumps from the refueling water storage tank was exempted in accordance with the clarification provided by IWC-1220(a) of the 1977/ Summer 1979 Edition / Addenda of ASME Section XI. Piping from the boric acid storage tank to the safety injection pump suction was also exempted in accordance with IWC-1220(a) .

6.0 INSPECTION REQUIREMENTS 6.1 For all components, except for the extGnt of examination of Class 1 and 2 piping welds, the following tables from the 1977/ Summer 1979 Edition / Addenda of ASME Section XI was utilized.

6.1.1 Class 1 - Table IWB-2500-1 6 .1. 2 Class 2 - Table IWC-2500-1 6.1.3 Class 3 - Table IWD-2500-1 6.1.4 Class 1, 2, and 3 supports - Table IWF-2500-1 6.2 For the extent of examination of Class 1 and 2 piping welds, the following tables from the 1974/ Summer 1975 Edition / Addenda of ASME Section XI will be utilized.

6.2.1 Class 1 - Table IWB-2500, Category BJ, and Table IWB-2600.

6.2.2 Class 2 - Table IWC-2520, Categories CF and CG, and Table IWC-2600.

6.3 The reactor coolant pump flywheel will be examined as specified in our 02/23/E2 letter from C. W. Fay to H. R. Denton. The inspection method and frequ.-ncy meet the intent of Regulatory Guide 1.14.

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.s Attachment 1

, Pago 4 7.0 RELIEF REQUESTS .

Those items for which Code .elief is requested can be found in Attachment 2.

8.0 CODE INTERPRETATIONS 8.1 Class 3 piping supports on piping greater than 4" diameter require examinations under Table IWF-2500-1. No examination will be performed on Class 3 piping supports on piping less than or equal to 4" diameter.

8.2 The regenerative heit exchanger. consists of three identical shells interconnected by piping as shown in Figure A-7. Each of the three identical shells are being considered similar in accordance with Note 3 of Table IWC-?500-1, Category CA. This applies to the Class 2 side only.

8.3 According to IWC-5210(a) (1), a system pressure test conducted during a system functional test of those Class 2 systems not required to operate during normal reactor operation is required every 40 months.

All Class 2 systems, according to IWC-5210(a) (2) , are required to have a 10-year hydrostatic pressure test.

9.0 ITEMS NOT COVERED BY THE LONG-TERM ISI PLAN 9.1 Snubber testing.is covered under Technical Specification 15.4.13.

l 9.2 Pump and valve testing, as specified under subsections IWP and IWV of t ASME Section XI, are handled under a separate pump and valve testing

program. This program was sul

snitted to the NRC on 02/10/81. The

! pump and valve program was written in accordance with the 1977/ Summer 1978 Edition / Addenda of ASME Section XI.

9.3 Steam generator tubing inspection is governed by Technical Specification 15.4.2. A versus ASME Section XI, as allowed by 10 CFR 50.53a(b) (2) (iii) .

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, n-I ATTACHMENT 2 12 FIT 2~ RELIEF REQUESTS

, , ' RELIEF REQUEST NO. DESCRIPTION.

Reactor vessel-interior exam.

.RR-2-1 .

'8R-2 -

Safety injection nozzle to safe end weld.'.

RR-2-3 Regenerative heat exchanger integrally welded supports.

~

RR-2-4 Reactor coolant pump integrally welded supports.

d. RR-2-5

~

Piping to penetration welds on the auxiliary coolant and safety injection systems.

RR-2-6 Reactor vessel integrally welded support attachment.

RR-2 -7 Class 3 integrally welded attachments.

RR-2-8 Reactor. coolant system' safety injection piping.

RR-2-9 Safety injection-reducer to safe-end weld.

i

~ DRAWINGS.

i FIGURE NO. DESCRIPTION ,

! 1 Regenerative heat exchanger' integrally welded attachment.

2 Containment wall penetration detaf1.

3 Reactor vessel internals.

A-1 Reactor pressure vessel.

A-7 Regenerative heat' exchanger.

A-8 Reactor coolant pump supports.

11 B-10 AC-10-RHR-2001 B-14 AC-10-RHR-2006~

B SIS 6-SI-2008 A-28 .

RC-6-SI-2001 and RC-4-SI-2001 A-32 RC-6-SI-2002 and RC-4-SI-2002 f

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,_ RR-2 '

COMPONENT

-Reactor pressure vessel.

EXAM AREA' Reactor vessel interior surfaces.

ISOMETRIC OR COMPONENT DRAWING None.

ASME SECTION XI CATEGORY

- BN1 ASME SECTION XI ITEM NUMBER B13.10 ,

ASME SECTION XI EXAMINATION REQUIREMENT A visual examination (VT-3) is required every 3 years of the accessible areas of the vessel interior surfaces'during a normal refueling outage.

ALTERNATIVE EXAMINATION A visual examination (VT-3) of the reactor vessel interior will be performed when the core barrel is removed but not at a frequency greater than that specified in the Code. The core barrel will not be removed

, specifically for this examination.

< REASON FOR LIMITATION As can be seen from Figure .3, only a small portion of the reactor vessel interior surfaces are accessible with the core barrel in place. There is approximately 10" of the vessel interior surface accessible from the top of the core barrel flange to the reactor vessel flange during a normal refueling outage. A meaningful examination cannot be performed unless the

• - core barrel is removed. Removal of the core barrel requires a complete defueling of the reactor and significant ALARA impacts including exposure and contamination problems.

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• 3 Attochment 2 Page 3 r

RR-2-2 COMPONENT IDENTIFICATION Reactor pressure vessel safety injection nozzles (2 nozzles).

EXAMINATION AREA Safe end-to-nozzle weld

l. RC-4-SI-2001-26

2. RC-4-SI-2002-29

~ ISOMETRIC OR COMPONENT DRAWING A-1 ASME SECTION XI CATEGORY BF ASME SECTION XI ITEM NUMBER B5.10 4

ASME SECTION XI EXAMINATION REQUIREMENT A surface and volumetric exam of each nozzle-to-safe end weld every 10 years.

ALTERNATIVE EXAMINATION A volumetric examination of each nozzle-to-safe end weld every 10 years.

This examination is performed from the ID of the nozzle using mechanized equipment.

AE.ASON FOR LIMITATION A surface examination is not possible due to the inaccessibility of this area. These welds are located between the vessel and biological shield wall.

', katochment 2 Page 4 RR-2-3 COMPONENT IDENTIFICATION Regenerative heat exchanger.

-EXAMINA. TON AREA Integrally welded supports (3) . .

ISOMETRIC OR COMPONENT DRAWING A-7 ASME SECTION XI CATEGORY BH ASME RECTION XI ITEM NUMBER B8.40

'ASME SECTION XI EXAMINATION REQUIREMENT A volumetric examination of each integrally welded attachment every 10 years,.

ALTERNATIVE EXAMINATION A visual examination (VT-3) of each integrally welded attachment every 10 years.

REASON F0R LIMITATION As can be seen in Figure 1, the integrally welded attachment is attached to the regenerative heat exchanger shell by an intermittent fillet weld.

These welds do not lend themselves to a meaningful ultrasonic, double wall radiographic, or alternative surface examination.

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Attrchment 2 Page 5 RR-2-4

, COMPONENT

' Reactor coolant pumps A and B.

EXAMINATION AREA Integrally welded support lugs (3).

ISOMETRIC OR CCMPONENT DRAWING A-8 ASME SECTION XI CATECORY BKl ASME SECTION XI~ ITEM NUMBER B10.20

- ASME SECTION XI EXAMINATION REQUIREMENT A volumetric examination of each integrally welded lug.is required every 10 years.

ALTERNATIVE EXAMINATION A visual examination (VT-3) of each integrally welded attachment will be performed every 10 years.

REASON FOR LIMITATION

' Ultrasonic ar.3 liquid penetrant examination of this area is not feasible due to the attenuative properties and surface roughness of the cast stainless steel.

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.,- Attichment 2 Page 6 RR-2-5 .

COMPONENT IDENTIFICATION Auxiliary coolant and safety injection systems.

_ EXAMINATION AREA

Piping-to-penetration cap welds.

1. AC-10-RHR-2001-24

2. AC-10-RHR-2006-18

3. SIS-6-SI-2008-19 ISOMETRIC OR COMPONENT DRAWING B-10,.B-14, B-29 ASME SECTION XI CATEGORY =

CF ASME SECTION XI ITEM NUMBER C5.ll ASKE SECTION XI EXAMINATION REQUIREMENT A surface examination of these welds is required.

ALTERNATIVE EXAMINATION

.No exam practical.

REASON FOR LIMITATION As shown in Figure 2, the welds of concern are totally enclosed within

'the containment penetration and are inaccessible.

4

. Attachment a Page 7 RR-2-6 COMPONENT Reactor pressure vessel.

EXAM AREA Reactor vessel integrally welded support attachment (2 attachments).

ISOMETRIC OR COMPONENT DRAWING None.

ASME SECTION XI CATEGORY

- BH ASME SECTION XI ITEM NO.

B8.10.

ASME SECTION XI EXAMINATION REQUIREMENT A volumetric or surface exam of each integrally welded attachment is required every ten years. According to Table IWB-2412-1, inspection Program B, the examination should be allocated as follows.

Inspection Period, Minimum Exams Maximum Exams Years Into Completed % Completed %

Interval 3 16 34 7 50 67 10 100 100 For two integrally welded attachments, one should be examined in the

' first or second period and one should be examined in the third period.

ALTERNATIVE EXAMINATION 4

A volumtric examination of both integrally welded attaclunents will be done once every ten years when the core barrel is removed from the reactor vessel.

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Attachment 2 Page 8 REASON FOR LIMITATION The vessel to integral attachment weld is examined from the inside surface of the reactor. vessel using a remote operated, mechanized, ultrasonic test

. device. To perform this exam,-the core barrel must be removed to permit access to the vessel wall'for ultrasonic examination. There is not access to the supports from the outside since they are located between the vessel and biological shield wall.

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Attachment 2 Page 9

, RR-2-7 COMPONENT Clans 3 integrally welded attachments.

EXAM AREA

' Component supports.

ISOMETRIC OR COMPONENT DRAWING None.

ASME SECTION XI CATECORY ,

DA, DB, DC ASME SECTION XI ITEM NUMER Dl.20 through Dl.60 D2.20 through D2.60 D3.20 through. D3.60 ASME SECTION XI EXAMINATION REQUIREMENT A visual examination (VT-3) is requir9d each inspection period of the integral. attachment including those attachments for c:mponents exceeding 4" nominal pipe size, whose strnettital integrity is relied upon to with-stand design loads when the system function is required.

ALTERNATE EXAMINATION l-A visual examination (VT-3) will be performed each inspection interval of the integral attachment including those attachments for components exceeding 4" nominal pipe size, whose structural integrity is relied upon to withstand design loads when the system function is required.

I REASON FOR LIMITATION

!- In the ASME Code Section XI 1977/S79 edition / addenda, the frequency of Class 3 support (section IWF) exams and their integrally welded attachments (section IWD) exams are as follows:

Exam Area Table _ Category _ Frequency Supports IWF-2500-1 FA, FB, FC Once per interva)

Integrally welded attachments IWD-2500-1 DA, DB, DC Once per period l

Attachment 2 Page 10

- The ext.mination requirements of Class 3 supports.and their integrally welded attachments are not consistent in Sections IWD and IWF of the 1977/S79 edition / addenda of ASME Section XI. The 1980/W80 edition / addenda revised the examination frequency requirements of the integrally welded attachments to once every interval. Based on the revision in the 1980/W80 edition /

addenda of ASME Section XI, and the inconsistency between sections IWD and IWF in the 1977/S79 edition / addenda of ASME Section XI, PBNP requests to examine Class 3 supports and their integrally wolded attachments at a frequency of once per interval.

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Attachment 2 Page 11 RR-2-8 COMPONENT IIANTICICATION Reactor coolant systems safety injection piping.

EXAMINATION AREA

1. 'RC-6-SI-2001-23A

2. RC-6-SI-2001-24

3. RC-6-SI-2002-26A

4. 'RC-6-SI-2002-27 I_SOMETRIC OR COMPONENT DRAWING

1. A-28

! 2. A-32 ASME SECTION XI CATEGORY BJ ASME SECTION XI ITEM NUMBER B9.11 ASME SECTION XI EXAMINATION REQUIREMENT A surface and volumetric exam of 25% of the circumferential joints each inspection interval (10 years) to accomplish 100s of the circumferential l joints over the service life of the plant.

ALTERNA2IVE EXAMINATION 1

l A surface and volunetric exam of 25% of the accessible circumferential joints each inspection interval (10 years) to accomplish 100% of the accessible circumferential joints over the service life of the plant will be performed.

The above examination areas are not accessible, leaving the number of accessible ASME Section XI, Category B-J circunferential welds at 92% of the

- total number of RC/AC-6-SI-2001 piping circumferential welds and 93% of the total number of RC/AC-6-SI-2002 piping circumferential welds.

REASON FOR LIMITATION

!~ A surface and volumetric exam from the exter.'ar of the piping is not possible due to the inaccessibility of this area. These welds are located between the reactor pressure vessel and biological shield wall. A volumetric examination l

from the . interior of the piping by mechanized equipment is also not possible due'to the pipe size and configuration.

Attachment 2 Page 12 RR-2-9 COMPONENT IDENTIFICATION Safety injection reducer to safe-end weld.

EXAMINATION ~ AREA

1. RC-4-SI-2001-25

2. RC-4-SI-2002-28 ISOMETRIC OR COMPONENT' DRAWING

1. A-28

2. A-32 ASME SECfION XI' CATEGORY BJ A3ME SECTION XI ITEM NUMBER 39.11 ASME SECTION XI EXAMINATION REQUIREMENT A surface and volumetric examination of 25% of the circumferentill joints each inspection interval (10 years).

ALTERNATIVE EXAMINATION A volumetric examination of each reducer to safe-end weld every 10 years when the associated nozzle to safe-end weld is performed. This examination is performed from the inside diameter of the safe-end using mechanized equipment.

REASON FOR LIMITATION A surface examination is not possible due to the inaccessibility of this area. These welris are located between the vessel and biological shield wall.

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ATTACHMENT 3 UNIT 1 RELIEF REQUESTS RELIEF REQUEST NO. DESCRIPTION RR-1-6 Reactor vessel integrally welded attachments.

RR-1-7 Class 3 integrally welded attaclanents.

RR-1-8 Reactor coolant syst m safety injection piping.

RR-1-9 Safety injection reducer to safe-end weld.

DRAWINGS FIGURE NO. DESCRIPTION A-26 RC-4-SI-1001, RC-6-SI-1001, AC-6-SI-1001 and SIS-2-SI-1001 A-29 RC-6-SI-1002, AC-6-SI-1002, RC-4-SI-1002, and SIS-2-SI-1003 e

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Attachment 3 Page 2 RR-1-6 COMPONENT Reactor pressure vessel.

t EXAM AREA Reactor vessel integrally welded support attachment (2 attachments).

ISOMETRIC OR COMPONENT' DRAWING .

None.-

ASME SECTION~XI CATEGORY BH

- ASME SECTION XI ITEM NO.

Bs.10 ASME SECTION XI' EXAMINATION' REQUIREMENT A' volumetric or surface exam of each integrally welded attachment is required every ten years. According to Table IWB-2412-1, Inspection Program B, the examination should be allocated as follows.

Inspection Period, Minimum Exams Maximum Exams Years into Interval Completed % Completed %

3 16 34 7 50 67 i 10 100 100 i

For two integrally welded attachments, one should be examined in the first or second period md one should be examined in the third period.

ALTERNATIVE EXAMINATION A vi.lumetric examination.of both-integrally welded attachments will be done once every ten years when the core barrel is removed from the reactor vessel.

REASON FOR L*.MITATION The vessel to integral attachment weld is examined from the inside surface of the reacter vessel using a remoto operated, mechaniszed, ultrasonic test device.

To perform this exam the core barrel must be removed to permit' access to the vessel wall for ultrasonic examination. There is no access to the supports from the outside since they are located between the vessel and biological shield wall.

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Attachment 3 Page 3 RR-1-7' COMPONENT Class 3 integrally welded attachments.

EXAM AREA Component sigports.

ISOMETRIC OR COMPONENT DRAWING None.

ASME Sm'"fION XI CATEGORY DA, DB, DC ASME SECTION'XI ITEM ~ NUMBER Dl.20 through D1.60 D2.20 through-D2.60 D3.20 through D3.60

~ASME SECTION XI~ EXAMINATION REQUIREMENT A visual exarination (VT-3) is required each inspection period of the integral attachmentrincluding those attachments for components exceeding 4" nominal-pipe size, whose structural integrity is relied upon to with-stand design loads when the system function is required.

ALTERNTl"E EXAMINATION A visual examination (VT-3) will be performed each inspection interval of the integral attachment, including those attachments for components exceeding 4" nominal pipe size whose structural integrity is relied upon to withstand

. design loads when the. system function is required.

REASON FOR LIMITATION In the ASME Code Section XI 1977/S79 edition / addenda, the frequency of Class 3 support (Sectic,n IWF) exams and their integrally welded attachments (Section IWD) exams are as follows:

g Area- Table Category Frequency Supports IWF-2500-1 FA, FB, FC Once per interval Integrally welded IWD-2500-1 DA, DB, DC Once per period

'Attachsents

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Attachment 3 Page 4 The examination requirements of Class 3 sugyorts and their integrally welded attachments are not consistent in Sections.IWD and IWF of the 1977/S79 edition /

addenda of ASME Section XI. The 1980/W80 edition / addenda revised the examination frequency requirements of the integrally welded attachments to once every interval. Based en the revision in the 1980/W80 edition / addenda of ASME Section XI, and the inconsistency between Sections IWD and IWF in the 1977/S79 edition / addenda of ASME-Section XI, PBNP requests to examine Class 3 supports and their integrally welded attachmente at a frequency of once per interval.

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4 Attachment 3 Page-5 RR-1 -8 COMPONENT IDENTIFICATION-Reactor coolant system safety injaction piping.-

EXAMINATION AREA

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^ 2. RC-6-SI-1001-29

3. RC-6-SI-1001-30

4. RC-6-SI-1001 31 t

5. RC-6-CI-1002-15

- 6.~ RC-6-SI-1002-16

7. RC-6-SI-1002-17 ISOMETRIC OR COMPONENT DRAWING

'l. A-26'

- 2. A-29 ASME SECTION XI CATEGORY BJ-ASME SECTION XI' ITEM' NUMBER i

B9.ll ASME SECTION XI EXAMINATION REQUIREMEfff A surface and volumetric . exam of 25% of the circumferential joints each inspection interval (10 years) = to accomplish 100% of the circumferential

. joints over the service life of the plant.

ALTERNATIVE EXAMINATION A' surface and volumetric exam of.25% of the accessible circumferential joints

.each inspection interval (10 years) to accomplish 100% of the accessible circumferential joints over the service life of the plant will be performed.

~ The above examination areas are not accessible, leaving the number of

. accessible ASME Section XIi Category.BJ circumferential welds at 88% of the

- total number-of BC-6-SI-1001 piping circumferential' welds and 84% of the j e /g(

total number of BC-6-SI-1002 piping circumferential welds. p

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Page 6 REASON FOR LIMITATION A' surface and voltsnetric exam from the exterior of the piping i3 not -

possible due to the inaccessibility of this area. These welds are located

.between-the. reactor pressure vessel and biological shield wall. A volumetric _ examination from ths interior of the piping by mechanized equipment is also not possible due to the pipe size and configuration.

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.Page 7 RR-1-9 l l

COMPONENT IDENTIFICATION' i-Safety injection reducer-to safe-end weld.

EXAMINATION AREA 1.- RC-4-SI-1001-32'-

2 . -- RC-4-SI-1002-18 ISOMETRIC OR COMPONENT DRAWINGS

1. 7-26 3

2. A-29 ASME SECTIOti XI CATEGORY BJ ASME SECTION XI ITEM NUMBER

• B9.11-

, ASME SECTION XI' EXAMINATION REQUIREMENT A surface and. volumetric examination of 25% of the circumferential joints each inspection interval ' (10 years) .

ALTERNATIVE EXAMINATION A volumetric examination of each reducer.to safe-end weld every 10 years r

when the associated nozzle to safe-end weld is performed. This examination is performed from the inside diameter of the safe-end using mechanized i equipment.

REASON FOR LIMITATION 4

A surface examination is not possible,due to '.he inaccessibility of this r -area. These welds are located between the vessel and biological shield wall.

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