ML20063B853

Requests NRC Approval of Encl Inservice Insp Relief Requests for Second 10-yr Insp.Approval of Tech Spec Change Request 42 Necessary for Adoption of ASME Section XI as Governing Document for Inservice Insp

ML20063B853
Person / Time
Site:	Point Beach
Issue date:	08/20/1982
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Clark R, Harold Denton Office of Nuclear Reactor Regulation
References
TAC-48762, NUDOCS 8208260240
Download: ML20063B853 (21)

Text

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i Msconsin Electnc eomcoww 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE, WI 53201 August 20, 1982 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attent' ion: Mr. R. A. Clark, Chief Operating Reactors, Branch 3 Gentlemen:

DOCKET NO. 50-266 INSERVICE INSPECTION PLAN POINT BEACH NUCLEAR PLANT, UNIT 1 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 1. In a letter dated October 6, 1981, Licensee notified the NRC of its plans to develop the Unit 1 long-term inservice inspection plan.

In Mr. Colburn's letter dated October 21, 1981, the NRC staff indicated that these plans were acceptable. The Unit 1 inservice inspection plan has been developed and approved by the Licensee.

As you know, due to the vintage of Point Beach Nuclear Plant, Unit 1 (the unit went into commercial service on December 21, 1970), the piping systems were designed prior to the concepts of Code Class 1, 2, and 3 and Safety Class 1, 2, and 3 systems now being used industry-wide by newer plants.. The pressure vessels at Point Beach Nuclear Plant under the scope of the Unit 1 inservice inspection plan were designed and constructed to Class A, B, or C standards of ASME Section III. The piping systems were designed and constructed in accordance with USAS B31.1. The design criteria utilized for Point Beach Nuclear Plant, Unit 1 were derived from the general design criteria of 10 CFR 50 as clarified in the Point Beach Nuclear Plant Final Safety Analysis Report.

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Mr. H. R. Denton August 20, 1982 l As stated in Paragraphs 5.2.1 and 5.2.2 of Attachment 1, Regulatory Guide 1.26 and ANSI Standard N18.2 were utilized as guidance to determine Code Class 1, 2, and 3 boundaries for inservice inspection purposes. Since Point Beach Nuclear Plant, Unit 1 was not originally designed to the specific requirements of Regulatory Guide 1.26 and ANSI Standard N18.2, the Code Class boundaries established for inservice inspection purposes meet the intent of the criteria 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 document is the foundation for the design and safety analysis of the unit. Attachment 1 provides a description of the Unit 1 inservice inspection second interval plan.

Relief requests for specific ASME Section XI Code requirements are provided in Attachment 2. As the second inspection interval progresses and other problem areas are identified, additional relief requests may have to be submitted.

As stated in Mr. R. A. Clark's letter dated March 12, 1982, the NRC's review of Technical Specification Change Request No. 42 is expected to be complete by August 1982. In order for Wisconsin Electric to fully implement the updated Unit 1 long-term inservice inspection plan, approval of Technical Specification Change Request No. 42 is necessary since this change request concerns adoption of ASME Section XI as the governing document for inservice inspection at Point Beach Nuclear Plant. We trust your review and approval of Change Request No. 42 is proceeding on schedule.

We hereby request NRC's acceptance of the ASME Section XI relief requests presented in Attachment 2 so our 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, Assistant Vice President C. W. Fay Attachments Copy to NRC Resident Inspector

ATTACHMENT 1 POINT BEACH NUCLEAR PLANT UNIT 1 ISI LONG-TERM PLAN (LTP)

DEVELOPMENT BASIS 1.0 PURPOSE To present a summary of the Unit 1 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 ASME Section XI.

2.2 The LTP provides a planning schedule of when a given examination will be performed during the second 10-year interval.

i 2.3 The LTP will provide an updated status of all required ISI.

3.0 APPLICABLE CODE EDITION / ADDENDA 3.1 All ISI work will be conducted in accordance with the 1977/ Summer 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 piping 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 examination of Class 1 and 2 piping welds will be in accordance with the 1977/ Summer 1979 Edition / Addenda of ASME Section XI.

4.0 BASIS FOR CODE SELECTION l

4.1 10 CFR 50 l 4.1.1 In accordance with 10 CFR 50.55a(g)(4)(ii), inservice exami-t nation of components and system pressure tests conducted during the second and successive 10 fear 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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Attachm:nt 1 Page 2 4.1.3 The 1974/ Summer 1975 Edition / Addenda of ASME Sectio XI must be utilized to determine the extent of examination of Class 2 piping welds in the residual heat removal and the safety 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 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 5.2.1 Regulatory Guide 1.26 (Revision 2), " Quality Group Classifi-cations and Standards 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. Article 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/ Summer 1975 Edition /

Addenda of ASME Section XI for piping welds.

Attachm:nt 1 Page 3 5.3.2 Class 2

a. Article IWC-1220 of the 1977/ Summer 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 NRC of exemption criteria IWC-1220(c) of the 1974/ Gummer 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 cooling function, pro-vided 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 INPECTION REQUIREMENTS 6.1 For all components, except for the extent of examination of Class 1 and 2 piping welds, the following tables from the 1977/ Summer 1979 Edition / Addenda of ASME Sectin 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 EJ, 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 02/23/82 letter from C. W. Fay to H. R. Denton. The inspection method and frequency meet the intent of Regulatory Guide 1.14.

Attachm:nt 1 Page 4 7.0 RELIEF REQUESTS Those items for which Code relief 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 examination 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 heat 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-2500-1, 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.

9.2 Pump and valve testing, as specified under subsections IWP and IWV of ASME Section XI, are handled under a separate pump and valve testing program. This program was submitted 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 Specifi-cation 15.4.2.A versus ASME Section XI, as allowed by 10 CFR 50.55a(b)(2)(iii).

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ATTACHMENT 2 UNIT 1 RELIEF REQUESTS RELIEF REQUEST NO. DESCRIPTION RR-1-1 Reactor vessel interior e::am.

RR-1-2 Safety injection nozzle to safe end weld.

RR-1-3 Regenerative heat exchanger integrally welded supports. .

RR-1-4 Reactor coolant pump integrally welded supports. ,

RR-1-5 Piping to penetration welds on the auxiliary coolant and safety injection systems.

DRAWINGS l

FIGURE NO. DESCRIPTION 1 Regenerative heat exchanger integrally welded attachment.

t i 2 Containment wall penetration detail.

3 Reactor vessel internals.

A-1 Reactor pressure vessel.

A-7 Regenerative heat exchanger.

A-8 Reactor coolant pump supports.

B-10 AC-10-RHR-1001 B-14 AC-10-RHR-1006 B-29 SIS-6-SI-1008 l

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Attachment 2

.Page 2 RR-1-1 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 4

B13.10 ASME SECTION XI EXAMINATION REQUIREMENT

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

l ALTERNATIVE EXAMINATION A visualexamination (VT-3) of the reactor vessel interio2 will be performed when the core barrel is removed but not at a frequency greater i

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 j ALARA impacts including exposure and contamination problems.

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Attachment 2 Page 3 RR-1-2 COMPONENT IDENTIFICATION Reactor pressure vessel safety injection nozzles (2 nozzles) .

EXAMINATION AREA Safe end-to-nozzle-weld

1. RC-4-SI-1001-33

2. RC-4-SI-1001-19 ISOMETRIC OR COMPONENT DRAWING 1 A-1 ASME SECTION XI CATEGORY BF ASME SECTION XI ITEM NUMBER B5.10 ASME SECTION XI EXAMINATION REQUIREMENT i

A surface and volumetric exam of each nozzle-to-safe end weld every 10 years.

f ALTERNATIVE EXAMINATION l 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.

REASON 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.

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Attachment 2 Page 4 RR-1-3

, COMPONENT IDENTIFICATION Regenerative heat exchanger.

EXAMINATION AREA

! Integrally welded supports (3).

ISOMETRIC OR COMPONENT DRAWING A-7 ASME SECTION XI CATEGORY Bl!

ASME XECTION 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 FOR 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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c Attachment 2 Page 5 i RR-1-4 COMPONENT Reactor coolant pump.

EXAMINATION AREA Integrally welded support lugs (3).

ISOMETRIC OR COMPONENT DRAWING

, A-8 ASME SECTION XI CATEGORY S

BK1 ASME SECTION XI ITEM NUMBER i

B10.20 ASME SECTION XI EXAMINATION REQUIREMENT A volumetric examination of each integrally welded lug is required i 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 and 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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t Attachment 2 Page 6 RR-1-5 COMPONENT IDENTIFICATION Auxiliary coolant and safety injection systems.

3 EXAMINATION AREA Piping-to-penetration cap welds

1. AC-10-RHR-1001-25

2. AC-10-RHR-1006-25

3. SIS-6-SI-1008-18 i

ISOMETRIC OR COMPONENT DRAWING B-10, B-14, B-29 ASME SECTION XI CATEGORY

! CF ASME SECTION XI ITEM NUMBER C5.11 ASME SECTION XI EXAMINATION REQUIREMENT A surface examination of these welds is required.

ALTERNATIVE EXAMINATION i

No exam practical.

REASON FOR LIMITATION i As shown in Figure 2, the welds of concern are totally enclosed within l the containment penetration and are inaccessible.

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