distribution: rdiggs aslabaslab gray file +4 whazelton rci limberg the commission has issued the...

Docket Nos. 50-266 and 50-301

Mr. C. W. Fay Assistant Vice President Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53201

Dear Mr. Fay:

DISTRIBUTION: LV11'cket File

NRC PDR L PDR NSIC ORB#3 Rdg DEisenhut PMKreutzer-3 RAClark TColburn OELD SECY I&E-2 TBranhart-8 LSchneider DBrinkman ACRS-1O OPA-CMiles

RDiggs ASLAB Gray File +4 WHazelton RCi limberg

The Commission has issued the enclosed Amendment No.6 3 to Facility Operating License No. DPR-24 and Amendment No. 68 to Facility Operating License No. DPR-27 for the Point Beach Nuclear Plant, Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Technical Specifications in partial response to your applications transmitted by letters dated February 17, 1977 and November 27, 1978 for Units I and 2 respectively.

These amendments revise the language for the Technical Specifications relating to inservice inspection requirements of safety class components to conform with the Codes and Standards Rule, 10 CFR 50.55a. This rule requires in part that inservice inspection of ASME Code Class 1, 2 and 3 components be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda except where specific written relief is granted by the Nuclear Regulatory Commission. The NRC staff provided specific guidance on the Technical Specification language recommended to conform with the above rule in a letter dated April 26, 1976.

Your applications for amendments dated February 17, 1977 and November 27, 1978 were responsive to the recommendations contained in our April 26, 1976 letter. Additionally, In those applications and another application for amendments dated December 8, 1977 you requested deletion of certain Point Beach Unit 1 and 2 Technical Specifications relating to the inservice testing requirements for ASME Code Class 1, 2 and 3 pumps and valves citing that the specific testing requirements for these pumps and valves were now superseded by Section XI of the ASME code. As discussed with members of your staff by telephone, those Technical Specification change requests relating to inservice inspection of pumps and valves will be addressed in separate correspondence upon completion of the NRC staff's review of the Point Beach Nuclear Plant Units 1 and 2 InservIce Inspection Program for Pumps and Valves (latest revision dated February 10, 1981).

As further discussed with members of your staff, the specific language in your Technical Specification change request did not quite conform with the iuidance provided in the NRC staff's April 26. 1976 letter.Additional wording exisled in your p1

OFFICE ........................

8210040414 8 a PDR ADOCK 05

NRC FORM 318 (10-80) NRCM 0240

20831 000266 I P- DR_.

-, I \.............

-oposed TechnIcal Spec fications ........................ ........................ ........................ ........................

........................ ........................ ........................ ........................

........................ ........................ ........................ ........................ t

OFFICIAL RECORD COPY USOPO: 1981-33�-96OOFICA REOR COP USGPO. 1981--335-960

J - • J Ir•ll•Ljj .....

I

Mr. C. W. Fay - 2

for Point Beach Units I and 2 to allow re-examination of components In a "mutually agreeable" manner should the NRC not agree with your documentation and proposed alternative examinations. The NRC staff feels this additional wording to be unnecessary and ambiguous and we have not approved its incorporation as an amendment to the Point Beach Unit 1 and 2 Technical Specifications. 10 CFR Part 50.55a allows for relief requests to be made and granted where ASME Code Section XI requirements prove to be impractical. These requests will be granted or denied based upon their individual merit and can be made at any time.

In addition to the above listed applications for amendments, by letters dated May 20, October 6, 1977, February 6, February 26, December 14, 1979, October 6, 1981, February 23, April 5, and April 14, 1982, you described your inservice inspection programs for Point Beach Units 1 and 2 and requested relief from certain requirements of ASME Code Section XI.

Our consultant, Science Applications, Inc. (SAI) has reviewed your requests for relief and has recommended relief should be granted from certain requirements of ASME Code Section XI. Based on SAI's review, summarized in the Technical Evaluation Report (TER) attached to the NRC staff's enclosed Safety Evaluation Report (SER), the NRC staff concludes that relief granted from certain examination and inspection requirements and the alternate methods discussed in the SER gives reasonable assurance of the piping component pressure boundary and support structural integrity. The NRC staff also concludes that granting relief where the code requirements are~impractical is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest considering the burden that could result upon Wisconsin Electric Power Company if these requirements were imposed on Point Beach Nuclear Plant Units 1 and 2.

Accordingly, pursuant to 10 CFR Section 50.55a(g)(6)(1) of the Commission's Regulations, Wisconsin Electric Power Company is granted relief from requirements for the following:

a. Volumetric examination of Unit 1 reactor vessel (RV) nozzle-tovessel welds and inside radiused sections at the frequency of two welds during the first period, one or two welds during the second period provided that a volumetric examination of all RV nozzle-to-vessel welds is done once every 10 years when the core barrel is removed.

b. Volumetric and surface examination of Unit 1 safety injection nozzle-to-safe end welds provided a volumetric examination of these welds and any problem welds are made during the shutdown if the surface examination of other reactor safe end welds indicates a surface flow.

DAT .......................E...................................................................................................................................... . . . . . . . . . . . . . . . .OFFICIAL RECORD COPY

I

NRC FORM 318 (10-80) NRCM 0240 USGPO: 1981--335-960

Mr. C. W. Fay

c. Volumetric examination of the Unit 1 and 2 reactor vessel integrally welded supports at the frequency of 25% during the first period, 25% during the second period and the remaining welds during the third period provided that a volumetric examination of 100% of the welds Is made when the core barrel Is removed during each 10 year Interval.

d. Volumetric examination of 10% of the Unit 1 and 2 regenerative heat exchanger integrally welded supports provided a visual examination Is conducted of these welds.

e. Volumetric examination of the Unit I and 2 reactor coolant pumps integrally welded supports provided a visual Jint dtpitis conducted oi? these welds.

f. Volumetric examination of the Unit 1 and 2 reactor coolant pumps casing welds provided these welds are examined in accordance with the 1977 Edition, Summer 1978 Addenda of the ASME Code Section XI.

Relief from the frequency to conduct visual inspection of the Unit I reactor vessel cladding patches is not necessary provided you update the Section XI ASME Code to the Summer 1978 Addenda for the category B-I-1 Items. Approval to update the requirements as stated above is required by the Commission and is hereby granted. The requirements for examining closure-head cladding and vessel cladding are deleted from the 1977 Edition with Addenda through Summer 1978.

Portions of the February 17, 1977 and November 27, 1978 Technical Specification change requests are being addressed in the safety evaluation as they relate to inservice Inspection of safety class components. The remaining portions of the above Technical Specification change requests and the December 8, 1977 Technical Specification change request relate to Inservice testing of pumps and valves and will be addressed in separate correspondance.

Copies of the Safety Evaluation and the Notice of Issuance are enclosed.

Sincerely,

Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Enclosures: 1. Amendment No. 3 to DPR-24 2. Amendment No. 6 8to DPR-27 3. Safety Evaluation 4. Notice of Issuance A/

FFICEL ........ ...... .......... . SUNAMEýP zer TCol burn/pn RCilfmbe.....elton &ClI.ar.rk L L G.Lain SUR AMEý 8~ ..... .... .. ........... .......8 /82 .. ...8/j .C/ 82 ..... 8,4 / .82

OFFICIAL RECORD COPY

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3.NR OM38(08)NC 024 USGO: 98133596

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NRC FORM 318 (10-80) NRCM 0240 USGPO: 1981--335-960

'11v UNITED STATES DISTRIBUTION: O NUCLEAR REGULATORY COMMISSION Docket File

WASHINGTON, D.C. 20555 ORB#3 Rdg PMKreutzer

Docket No. 50-266/50-301

Docketing and Service Section Office of the Secretary of the Commission

SUBJECT: WISCONSIN ELECTRIC POWER COMPANY, Point Beach Nuclear Plant Unit Nos. 1 and 2

Two signed originals of the Federal Register Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication. Additional conformed copies ( 12 ) of the Notice are enclosed for your use.

El Notice of Receipt of Application for Construction Permit(s) and Operating License(s).

El Notice of Receipt of Partial Application for Construction Permit(s) and Facility License(s): Time for Submission of Views on Antitrust Matters.

El Notice of Availability of Applicant's Environmental Report.

El Notice of Proposed Issuance of Amendment to Facility Operating License.

El Notice of Receipt of Application for Facility License(s); Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility License(s) and Notice of Opportunity for Hearing.

El Notice of Availability of NRC Draft/Final Environmental Statement.

E] Notice of Limited Work Authorization.

El Notice of Availability of Safety Evaluation Report.

El Notice of Issuance of Construction Permit(s).

El Notice of Issuance of Facility Operating License(s) or Amendment(s).

RI Other: Amendment Ns.. 63 a"d 6g.

Rpferpnryd idocuments have been provided PDR.

Divisio. of L en inja Office of Nuclear Reactor IRegulation

Enclosure: •As Stated

UR NAOR.P.m .102.

-... ... .

NRC FORM 102 7 -- 79

Wisconsin Electric Power Company

cc: Mr. Bruce Churchill, Esquire Mr. William Guldemond Shaw, Pittman, Potts and Trowbridge USNRC Resident Inspectors Office 1800 M Street, N. W. 6612 Nuclear Road Washington, D. C. 20036 Two Rivers, Wisconsin 54241

Joseph Mann Library 1516 Sixteenth Street Two Rivers, Wisconsin 54241

Mr. Glenn A. Reed, Manager Nuclear Operations Wisconsin Electric Power Company Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, Wisconsin 54241

Mr. Gordon Blaha Town Chairman Town of Two Creeks Route 3 Two Rivers, Wisconsin 54241

Ms. Kathleen M. Falk General Counsel Wisconsin's Environmental Decade 114 N. Carroll Street Madison, Wisconsin 53703

U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN: Regional Radiation

Representative 230 S. Dearborn Street Chicago, Illinois 60604

cc w/enclosure(s*) and incoming dtd: 2/17/77, 11/27/78

Chairman Public Service Commission of Wisconsin Hills Farms State Office Building Madison, Wisconsin 53702

Regional Administrator Nuclear Regulatory Commission, Region III Office of Executive Director for*Operations 799 Toosevelt Road Glen Ellyn, Illinois 60137

UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

WISCONSIN ELECTRIC POWER COMPANY

DOCKET NO. 50-266

POINT BEACH NUCLEAR PLANT, UNIT NO. 1

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 63 License No. DPR-24

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Wisconsin Electric Power Company (the licensee) dated February 17, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

B. The facility will the provisions of the Commission;

operate in conformity with the application, the Act, and the rules and regulations of

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

DESIGNATED ORIGINAL

Certified By

82160040420 820831 PDR ADOCK 05000266 P

PDR

-2-

2. Accordingly, the license is Specifications as indicated amendment, and paragraph 3.B No. DPR-24 is hereby amended

amended by changes to the Technical in the attachment to this license of Facility Operating License to read as follows:

(B) Technical Specifications

The Technical Specifications contained in Appendices

A and B, as revised through Amendment No. 63 , are

hereby incorporated in the license. The licensee shall

operate the facility in accordance with the Technical Specifications.

3. This license amendment is issuance.

effective 20 days from the date of its

FOR THE NUCLEAR REGULATORY COMMISSION


Attachment: Changes to the Technical

Specifications

Date of Issuance: August 31, 1982

ATTACHMENT TO LICENSE AMENDMENTS

AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. DPR-24


DOCKET NOS. 50-266 AND 50-301

Revise Appendix A as follows:

Remove Pages

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15.6.9-7

TABLE OF CONTENTS

Title

TECHNICAL SPECIFICATIONS AND BASES

Definitions

15.2.0 15.2.1 15.2.2 15.2.3

15.3 15.3.0 15.3.1 15.. 3 .:2 15.3.:3

15.3.4 15.3.5 15.3.6 15.3.7 15.3.8 15.3.9 15.3.10 15.3.11 15.3.12 15.3.13 15.3.14 15.3.15 15.3.16

15.4 15.4.1 15.4.2 15.4.3 15.4.4 15.4.5

15.4.6 15.4.7 15.4.8 15.4.9 15.4.10 15.4.11 15.4.12 15.4.13 15.4.14 15.4.15 15.4.16

15-i

Section

15

15.1

Safety Limits and Limiting Safety System Settings Safety Limit, Reactor Core Safety Limit, Reactor Coolant System Pressure Limiting Safety System Settings, Protective

Instrumentation

Limiting Conditions for Operation General Considerations Reactor Coolant System Chemical and Volume Control System Emergency Core Cooling System, Auxiliary Cooling

Systems, Air Recirculation Fan Coolers, and Containment Spray

Steam and Power Conversion System Instrumentation System Containment System Auxiliary Electrical Systems Refueling Effluent Releases Control Rod and Power Distribution Limits Movable In-Core Instrumentation Control Room Emergency Filtration Shock Suppressors (Snubbers) Fire Protection System Overpressure Mitigating System Reactor Coolant System Pressure Isolation Valves

Surveillance Requirements Operational Safety Review In-Service Inspection of Safety Class Components Primary System Testing Following Opening Containment Tests Emergency Core Cooling System and Containment

Cooling System Tests Emergency Power System Periodic Tests Main Steam Stop Valves Auxiliary Feedwater System Reactivity Anomalies Operational Environmental Monitoring Control Room Emergency Filtration Miscellaneous Radioactive Materials Sources Shock Suppressors (Snubbers) Surveillance of Auxiliary Building Crane Fire Protection System Reactor Coolant System Pressure Isolation Valves

Leakage Tests

Page

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15.4-1 15.4.1-1 15.4.2-1 15.4.3-1 15.4.4-1

15.4.5-1 15.4.6-1 15.4.7-1 15.4.8-1 15.4.9-1 15.4.10-1 15.4U11-i 15.4.12-1 15.4.13-1 15.4.14-1 15.4.15-1

15.4.16-1

unit 1 - OU jifX A 01 U1~$, 63 Unit 2 - OU0t Aý$tI10 68

I

15.4.2 IN-SERVICE INSPECTION OF SAFETY CLASS COMPONENTS

Applicability

Applies to in-service inspection of Safety Class Components.

Obiectives

To provide assurance of the continuing integrity of the safety class systems..

Specifications

A. Steam Generator Tube Inspection Requirements

1. Tube Inspection

Entry from the hot-leg side with examination from the point of entry completely around the U-bend to the top support of the cold-leg is considered a tube inspection.

2. Sample Selection and Testing

Selection and testing of steam generator tubes shall be made on the following basis:

(a) One steam generator of each unit shall be inspected during inservice inspection in accordance with the following requirements:

1. The inservice inspection may be limited to one steam generator on an alternating sequence basis. This _ examination shall include at least 6% of the tubes if the results of the first or a prior inspection indicate that both generators are performing in a comparable manner.

2. When both steam generators are required to be examined by Table 15.4.2-1 and if the condition of the tubes in one generator is found to be more severe than in the other steam generator of a unit, the steam generator sampling sequence at the subsequent inservice inspection shall be modified to examine the steam generator with the more severe condition.

(b) The minimum sample size, inspection result classification and the associated required action shall be in conformance with the requirements specified in Table 15.4.2-1. The results of each sampling examination of a steam generator shall be classified into the following three categories:

Point Beach Unit 1 15.4.2-1 Amendment No. 10, 63

Defect is an imperfection of such severity that it exceeds the minimum acceptable tube wall thickness of 50%. A tube containing a defect is defective.

Plugging limit is the imperfection depth beyond which the tube must be removed from service, because the tube may become defective prior to the next scheduled inspection. The plugging limit is 40% of the nominal tube wall thickness.

6. Corrective Measures

All tubes that leak or have degradation exceeding the plugging limit shall be plugged prior to return to power from a refueling or inservice inspection condition.*

7. Reports

(a) After each inservice examination, the number of tubes plugged in each steam generator shall be reported to the Commission as soon as practicable. . .

(b) The complete results of the steam generator tube inservice inspection shall be included in the Operating Report for the period in which the inspection was completed. In addition all results in Category C-3 of Table 15.4.2-1 shall be reported to the Commission prior to resumption of plant operation.

(c) Reports shall include:

1. Number and extent of tubes inspected

2. Location and percent of all thickness penetration for each indication

3. Identification of tubes plugged

(d) Reports required by Table 15.4.2-1 - Steam Generator Tube Inspection - shall provide the information required by Specification 15.4.2.A.7(b) and a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

B. In-Service Inspection of Safety Class Components Other than Steam Generator Tubes

1. Inservice inspection of ASME Code Class 1, Class 2 and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specific written relief is granted by the NRC, pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

*Point Beach Nuclear Plant Unit 1 may be operated at power with up to six tubes in one steam generator having degradation exceeding the plugging limit provided those tubes have been repaired by insertion of sleeves into the tubes to bridge the degraded or defective portion of the tube. The plugging limit is 35% of the nominal sleeve wall thickness for tubes that have been repaired by sleeving.

Point Beach Unit 1 15.4.2-1c Amendment No. Z0, Z0, 63

2. Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps

and valves shall be performed in accordance with Section XI of the

ASME Boiler and Pressure Vessel Code and applicable Addenda as

required by 10 CFR 50, Section 50.55a(g) modified by Section

50.55a(b), except where specified written relief is granted by the

NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

3. Containment isolation valves will be tested in accordance with

Technical Specification 15.4.4 instead of Section IWV-3420, Valve

Leak Rate Test.

Bases

The proposed inspection program is, where practical, in comp.Liance with the

recommendations of ASME Boiler and Pressure Vessel Code, Section XI, Summer

1971 Addenda. It must be recognized, however, that equipment and techniques

to perform the inspection are still in development. It is recognized, however,

that examinations in certain areas are necessary and therefore a schedule is

proposed that includes areas and frequencies that are believed practical at

this time for this reactor. In most areas scheduled for test, a detailed

pre-service mapping will be conducted using techniques which can be used for

post-operation inspections. The areas indicated for inspection represent those

of relatively high stress and therefore will serve to indicate potential

problems before significant flaws develop there or at other areas. As more

experience is gained in operation of pressurized-water reactors, the recommended

time schedule and location of inspection might be altered, or should new

techniques be developed, consideration will be given to incorporate these new

techniques into this inspection program.

The use of conventional non-destructive, direct visual and remote visual test

techniques can be applied to the inspection of all primary loop components

except for the reactor vessel. The reactor vessel presents special problems

because of the radiation levels and remote underwater accessibility to this

component. Because of these limitations on access.to the reactor vessel,

several steps have been incorporated into the design and manufacturing procedures

in preptution for non-destructive test techniques which may be available in the future.

The techniques for in-service inspection include the use of visual inspections,

volumetric (ultrasonic or radiographic) and surface (dye penetrant or magnetic

particle) testing of selected parts during refueling periods.

The intent of the inspection is the detection of flaws large enough to initiate

fast fracture and gross leakage prior to subsequent inspection. At this time

it is judged that such a flaw is substantially larger than 1/2 inch by 1 inch

which is the degree of detectability. The inspection method is designed to

detect flaws of this magniture.

(1) FSAR - Section 4.4

15.4.2-2 Amendment No. 63Point Beach Unit 1

3. Observed inadequacies in the implementation of admin- -..

istrative or procedural controls which threaten to cause

reduction of degree of redundancy, provided in reactor

protection systems or engineered safety feature systems.

4. Abnormal degradation of systems other than those specified

in 15.6.8.2.A.3 above designed to contain radioactive

material resulting from the fission process.

15.6.9.3 UNIQUE-REPORTING REQUIREMENTS

The following written reports shall be submitted to the Director,

Office of Nuclear Reactor Regulation, USNRC:

A. Each integrated leak test shall be the subject of a summary

technical report, including results of the local leak rate

tests and isolation valve leak rate tests since the last

report. The report shall include analysis and interpreta

tions of the results which demonstrate compliance with

specified leak rate limits.

B. Deleted

C. Submission of a report within 60 days after January 1 and

after July 1 each year for the six-month period or fraction

thereof, ending June 30 and December 31 containing:

Amendment No. 11, 6315.6.9-7Point Beach Unit 1

,o.x UNITED STATES 0 NUCLEAR REGULATORY COMMISSION

/ WASHINGTON, D. C. 20555


DOCKET NO. 50-301

POINT BEACH NUCLEAR PLANT, UNIT NO. 2

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 68 License No. DPR-27

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Wisconsin Electric Power Company (the licensee) dated November 27, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the healthand safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

DESIGNATED ORIGINAL

Ce'tifled By

-2-

2. Accordingly, the license is Specifications as indicated amendment, and paragraph 3.B No. DPR-27 is hereby amended

amended by changes to the Technical in the attachment to this license of Facility Operating License to read as follows:

(B) Technical Specifications

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 68 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective 20 days from the date of its issuance.



Attachment: Changes to the Technical

Specifications

Date of Issuance: August 31, 1982

ATTACHMENT TO LICENSE AMENDMENTS




Revise Appendix A as follows:

Remove Pages

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15.6.9-7

Section

15-i Unit 1 Unit 2 - C lit

Page

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15.1

0119 Z0Z,//9,.

15.2.0 15.2.1 15.2.2 15.2.3

15.3 15.3.0' 15.3.1 15.3.2 15.3.3

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15.4.6 15.4.7 15.4.8 15.4.9 15.4.10 15.4.11 15.4.12 15.4.13 15.4.14 15.4.15 15.4.16

TABLE OF CONTENTS

Title

TECHNICAL SPECIFICATIONS AND BASES

Definitions

Safety Limitsand Limiting Safety System Settings Safety Limit, Reacior Core Safety Limit, Reactor Coolant System Pressure Limiting Safety System Settings, Protective

Instrumentation

Limiting Conditions for Operation General Considerations Reactor Coolant System Chemical and Volume Control System Emergency Core Cooling System, Auxiliary Cool4ng

Systems, Air Recirculation Fan Coolers, and Containment Spray

Steam and Power Conversion System Instrumentation System Containment System Auxiliary Electrical Systems Refueling Effluent Releases Control Rod and Power Distribution Limits Movable In-Core Instrumentation Control Room Emergency Filtration Shock Suppressors (Snubbers) Fire Protection System Overpressure Mitigating System Reactor Coolant System Pressure Isolation Valves

Surveillance Requirements Operational Safety Review In-Service Inspection of Safety Class Components Primary System Testing Following Opening Containment Tests Emergency Core Cooling System and Containment

Cooling System Tests Emergency Power System Periodic Tests Main Steam Stop Valves Auxiliary Feedwater System Reactivity Anomalies Operational Environmental Monitoring Control Room Emergency Filtration Miscellaneous Radioactive Materials Sources Shock Suppressors (Snubbers) Surveillance of Auxiliary Building Crane Fire Protection System Reactor Coolant System Pressure Isolation Valves

Leakage Tests

15.1-1

15.2.1-1 15.2.1-1 15.2.2-1

15.2.3-1

15.3-0 15.3.0-1 15.3.1-1 15.3.2-1

15.3.3-1 15.3.4-1 15.3.5-1 15.3.6-1 15.3.7-1 15.3.8-1 15.3.9-1 15.3.10-1 15.3.11-1 15.3.12-1 15.3.13-1 15.3.14-1 15.3.15-1 15.3.16-1

15.4-1 15.4.1-1 15.4.2-1 15.4.3-1 15.4.4-1

15.4.5-1 15.4.6-1 15.4.7-1 15.4.8-1 15.4.9-1 15.4.10-1 15.4-. 11-1 15.4.12-1 .15.4.13-1 15.4.14-1 15.4.15-1

15.4.16-1

63 68

I

15.4.2 IN-SERVICE INSPECTION OF SAFETY CLASS COMPONENTS

Applicability

Applies to in-service inspection of Safety Class Components.

Objectives

To provide assurance of the continuing integrity of the safety class systems.

Specifications

A. Steam Generator Tube Inspection Requirements

1. Tube Inspection

Entry from the hot-leg side with examination from the point of entry completely around the U-bend to the top support of the cold-leg is considered a tube inspection.

2. Sample Selection and Testing

Selection and testing of steam generator tubes shall be made on the following basis:

(a) One steam generator of each unit shall be inspected during inservice inspection in accordance with the following requirements:

1. The inservice inspection may be limited to one steam generator on an alternating sequence basis. This examination shall include at least 6% of the tubes if the results of the first or a prior inspection indicate that both generators are performing in a comparable manner.

2. When both steam generators are required to be examined by Table 15.4.2-1 and if the condition of the tubes in one generator iw found to be more severe than in the other steam generator of a unit, the steam generator sampling sequence at the subsequent inservice inspection shall be modified to examine the steam generator with the more severe condition.

(b) The minimum sample size, inspection result classification and the associated required action shall be in conformance with the requirements specified in Table'15.4.2-1. The results of each sampling examination of a steam generator shall be classified into the following three categories:

Point Beach Unit 2 15.4.2-1 Amendment No. 11, 68

Defect is an imperfection of such severity that it exceeds the minimum acceptable tube wall thickness of 50%. A tube containing a defect is defective.

Plugging Limit is the imperfection depth beyond which the tube must be removed from service, because the tube may become defective prior to the next scheduled inspection. The plugging limit is 40% of the nominal tube wall thickness.

6. Corrective Measures

All tubes that leak or have degradation exceeding the plugging limit shall be plugged prior to return to power from a refueling or inservice inspection condition.

7. Reports

(a)>. After each inservice examination, the uumber of tubes plugged in each steam generator shall be reported to the Commission as soon as practicable.

(b) The complete results of the steam generator tube inservice inspection shall be included in the Operating Report for the period in which the inspection was completed. In addition all results in Category C-3 of Table 15.4.2-1 shall be reported to the Commission prior to resumption of plant operation.

(c) Reports shall include:

1. Number and extent of tubes inspected

2. Location and percent of all thickness penetration for each indication

3. Identification of tubes plugged

(d) Reports required by Table 15.4.2-1 - Steam Generator Tube Inspection - shall provide the information required by Specification 15.4.2.A.7(b). and a description of investigationsi conducted to determine cause of the tube-degradation and corrective measures taken to prevent recurrence.

B. In-Service Inspection of Safety Class Components Other Than Steam Generator Tubes

1. Inservice inspection of ASME Code Class 1, Class 2 and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specific written relief is granted by the NRC, pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

Amendment No. 12, 68Point Beach Unit 2 15.4.2-1c

2. Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specified written relief is granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

3. Containment isolation valves will be tested in accordance with Technical Specification 15.4.4 instead of Section IWV-3420, Valve Leak Rate Test.

Bases

The proposed inspection program is, where practical, in comB3.±ance with the recommendations of ASME Boiler and Pressure Vessel Code, Section XI, Summer 1971 Addenda. It must be recognized, however, that equipment and techniques to perform the inspection are still in development. It is recognized, however, that examinations in certain areas are necessary and therefore a schedule is proposed that includes areas and frequencies that are believed practical at this time for this reactor. In most areas scheduled for test, a detailed pre-service mapping will be conducted using techniques which can be used for post-operation inspections. The areas indicated for inspection represent those of relatively high stress and therefore will serve to indicate potential problems before significant flaws develop there or at other areas. As more experience is gained in operation of pressurized-water reactors, the recommended time schedule and location of inspection might be altered, or should new techniques be developed, consideration will be given to incorporate these new techniques into this inspection program.

The use of conventional non-destructive, direct visual and remote visual test techniques can be applied to the inspection of all primary loop componen-ts except for the reactor vessel. The reactor vessel presents special problems because of the radiation levels and remote underwater accessibility to this component. Because of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufacturing procedures in prepmjtion for non-destructive test techniques which may be available in the future.

The techniques for in-service inspection include the use of visual inspections, volumetric (ultrasonic or radiographic) and surface (dye penetrant or magnetic particle) testing of selected parts during refueling periods.

The intent of the inspection is the detection of flaws large enough to initiate fast fracture and gross leakage prior to subsequent inspection. At this. time it is judged that such a flaw is substantially larger than 1/2 inch by 1 inch which is the degree of detectability. The inspection method is designed to detect flaws of this magniture.

(1) FSAR - Section 4.4

15.4.2-2 Amendment No. 68Point Beach-Unit 2

3. Observed inadequacies in the implementation of admin

istrative or procedural controls which threaten to cause

reduction of degree of redundancy provided in reactor

protection systems or engineered safety feature systems.

4. Abnormal degradation of systems other than those specified

in 15.6.8.2.A.3 above designed to contain radioactive

material resulting from the fission process.

15.6.9.3 UNIQUE REPORTING REQUIREMENTS

The following written reports shall be submitted to the Director,

Office of Nuclear Reactor Regulation, USNRC:

A. Each integrated leak test shall be the subject of a summary

technical report, including results of the local leak rate

tests and isolation valve leak rate tests since the last

report. The report shall include analysis and interpreta

tions of the results which demonstrate compliance with

specified leak rate limits.

B. Deleted

C. Submission of a report within 60 days after January 1 and

after July 1 each year for the six-month period or fraction

thereof, ending June 30 and December 31 containing:

Amendment No. 1, 68

"i-

Point Beach Unit 2 15.6.9-7

'b0 UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

SUPPORTING AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. DPR-24

AND AMENDMENT NO. 68 TO FACILITY OPERATING LICENSE NO. DPR-27


POINT BEACH NUCLEAR-PLANT, UNIT NOS. 1 AND 2


Introduction

By letter dated April 26, 1976, the NRC outlined to Wisconsin Electric Power Company (licensee) the requirements of the newly enacted rule governing inservice inspection of safety class components, 10 CFR 50.55a. That letter also requested that the licensee update its Technical Specifications to conform to the new rule and, as allowed by the new rule, request relief and provide justification for those requirements, if any, felt to be impractical to perform for Point Beach Nuclear Plant, Units 1 and 2. The licensee responded with Technical Specification (TS) change requests by letters dated February 17, 1977 (Unit 1), December 8, 1977 (Units land 2) and November 27, 1978 (Unit 2).

Proposed TS 15.4.2.B for the Point Beach Nuclear Plants Units 1 and 2 states that inservice examination of ASME Code Glass 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by the 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission. Certain requirements of later editions and addenda of Section XI are impractical to perform on older plants because of the plants' design, component geometry, and materials of construction. Thus, 10 CFR 50.55a(g)(6)(i) authorizes the Commission to grant relief from those requirements upon making the necessary findings.

By letters dated May 20, 1977, October 6, 1977, February 6, 1979, February 26, 1979, December 14, 1979, October 6, 1981 and April 14, 1982, Wisconsin Electric Power Company submitted its inservice inspection program, revisions, or additional information related to requests for relief from certain Code requirements determined to be impractical to perform on the Point Beach facilities during the inspection interval. The inservice inspection programs are based on the requirements of the 1974 Edition through Summer 1975 Addenda of Section XI.

DESIGNATED ORIGINAL

Certified By__

2000425 82 0 63 PDR oDOCK& 05000266 PPDR

'��2

-2

Evaluation

Requests for relief from the requirements of Section.XI which have been determined to be impractical to perform have been reviewed.by the staff's contractor, Science Applications, Inc. The contractor's evaluations of the licensee's requests for relief and his recommendations are presented in the Technical Evaluation Report (TER) attached (Attachment 1). The staff has reviewed the TER and agrees with the evaluations and recommendations. A summary of the determinations made by the staff is presented in the following table:

Table 1 Class 1 Components

Licensee proposed Relief

IWB-2600 IWB-2500 System or Area to be Required alterna- request item no. exam. cat. component examined method tive exam. status

Nozzle-tovessel welds and inside radiused sections

Weld

Integrallywelded supports

Volumetric at frequency below: 1st period

2 welds 2nd perid

1 or 2 welds

3rd period remaining welds

Volumetric & surface at frequency in IWB-2411

Volumetric at frequency below: 1st period

25% 2nd period

25% 3rd period

remainder

Volumetric all nozzles once every 10 years when core barrel is removed _

Granted

Volumetric Granted only once every 10 years when core barrel is removed

Volumetric100% of weld when core barrel is removed during interval

.. Granted

B-D Reactor vessel nozzles (6)

B1.4 (Applies to Unit 1 only)'

B1.6 (Applies to Unit 1 only)

B1.12 (Units 1 & 2)

B-F Safety Injection nozzleto-safe end

B-H Reactor vessel

-3-

Table 1 (Continued)

Licensee proposed Relief

IWB-2600 IWB-2500 System or Area to be Required alterna- request item no. exam. cat. component examined method tive exam. status

B1.14 B-I-1 Reactor Cladding Visual at Visual - Update to (Applies vessel patches frequency IO%-when 1977 edit-on to Unit 1 below: core barrel through only) Ist period - is removed Summer 1978

25% addenda; 2nd period - relief nct

25% necessary 3rd period

remainder

B3.7 B-H Regenera- Integrally - Volumetric Visual Granted (Units 1 tive heat welded sup- (10% of & 2) exchanger ports weld)

B5.4 B-K-1 Reactor Integrally - Volumetric Visual Granted (Units 1 coolant welded & 2) pumps supports

B5.6 B-L-1 Reactor Pump Volumetric Examine Weld Granted (Units 1 coolant casing To 1977 578 & 2) pumps welds Section XI Code

Environmental Consideration

We have determined that granting relief from specific ASME Section XI

Code requirements does not authorize a change in effluent types or total

amounts nor an increase in power level and will not result in any signi

ficant environmental impact. Having made this determination, we have

further concluded that this is an action which is insignificant from

the standpoint of environmental impact and, pursuant to 10 CFR §51.5(d)(4),

that an environmental impact statement or negative declaration and environ

mental impact appraisal need not be prepared in connection with the granting

of this relief.

-4-

Conclusion

The NRC staff has reviewed portions of the February 17, 1977 and November 27, 1978 TS change requests and after making modifications to the wording as discussed with the licensee's staff, has found them acceptable. These proposed TS relating to inservice inspection of safety class com

ponents, as modified, conform to the language recommended by the NRC

staff's April 26, 1976 letter and are, therefore, acceptable.

Based on the review summarized, the staff concludes that relief granted

from the examination requirements and alternate methods imposed through this document give reasonable assurance of the piping, component pressure boundary, and support structural integrity, that granting relief where

the Code requirements are impractical is authorized by law and will not

endanger life or property, or the common defense and security, and is

otherwise in the public interest considering the burden that could result

if they were imposed on the facility.

We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in

the probability or consequences of an accident previously evaluated,

does not create the possibility of an accident of a type different from

any evaluated previously, and does not involve a significant reduction

in a margin of safety, the amendment does not involve a significant

hazards consideration, (2) there is reasonable assurance that the health

and safety of the public will not be endangered by operation in the

proposed manner, and (3) such activities will be conducted in compliance

with the Commission's regulations and the issuance of this amendment Will

not be inimical to the common defense and security or to the health and

safety of the public.

Date: August 31, 1982

Principal Contributors:

T. Colburn G. Johnson

SAI•'port No. 186-028-22

POINT BEACH NUCLEAR PLANT

UNITS 1 AND 2

INSERVICE INSPECTION PROGRAM

TECHNICAL EVALUATION REPORT

Submitted to:

U.S. Nuclear Regulatory Commission

Contract No. 03-82-096

Science Applications, Inc. McLean, Virginia 22102

June 15, 1982

ATTACHMENT I

# W ,

CONTENTS

INTRODUCTION ......... ..... ....... ........................ 1

I. CLASS 1COMPONENTS ......... ....... ..................... 4

A. Reactor Vessel ..... .............. .................. 4

1. Nozzle-to-Shell Welds and Nozzle Inside Radiused SSections (Unit 1); Category B-D, Item B1.4 .... ........ 4

2. Safety Injection Nozzle to Safe End Welds (Unit 1); Category B-F, Item B1.6 ........ ............ 6

3. Integrally Welded Supports (Unit 1); Category B-H, Item BI.12 ......... ............... .. .. ........ 8

4. Vessel Cladding (Unit 1); Category B-I-l, Item B1.14 . . 9

B-. Pressurizer (No relief requests)

C. Heat Exchangers and Steam Generators ...... ............. 11

1. Regenerative Heat Exchanger, Integrally Welded Supports (Units 1 & 2); Category B-H, Item B3.7 ... ..... 11

D. Piping Pressure Boundary (No relief requests)

E. Pump Pressure Boundary ....... ................... ... 12

1. Reactor Coolant Pumps, Integrally Welded Supports (Units 1 & 2); Category B-K-I, Item B5.4 ..... ......... 12

2. Reactor Coolant Pumps, Pump Casing Welds (Units 1

& 2); Category B-L-1, Item B5.6 ..... ............. ... 14

F. Valve Pressure Boundary (No relief requests)

II. CLASS 2 COMPONENTS (No relief requests)

III. CLASS 3 COMPONENTS (No relief requests)

IV. PRESSURE TESTS (No relief requests)

V. GE,4ERAL (No relief requests)

REFERENCES ....... ... ..................... . ........ 17

"* TE•ICAL EVALUATION REPORT

POINT BEACH NUCLEAR PLANT UNITS 1 & 2

INSERVICE INSPECTION PROGRAM

INTRODUCTIOON

The revision to 10 CFR 50.55a, published in February 1976, required that

Insqrvice Inspection (ISI) Programs be updated to meet the requirements (to

the extent practical) of the Edition and Addenda of Section XI of the American

Society of Mechanical Engineers Boiler and Pressure Vessel Code* incorporated

in the Regulation by reference in paragraph (b). This updating of the programs

was required to be done every forty months to reflect the new requirements of

the later editions of Section XI.

As specified in the February 1976 revision, for plants with-Operating

Licenses issued prior to-March 1, 1976, the Regulations became effective after

September 1, 1976 at the start of the next regular 40-month inspection period.

The initial inservice examinations conducted during the first 40-month period

were to comply with the requirements in editions of Section XI and addenda in

effect no more than six months prior to the date of start of facility commercial

operation.

The Regulation recognized that the requirements of the later editions and

addenda of the Section XI might not be practical to implement at facilities be

cause of limitations of design, geometry, and materials of construction of

components and systems. It therefore permitted determinations of impractical

examination or testing requirements to be evaluated. Relief from these require

ments could be granted provided health and safety of the public were not endangered

giving due consideration of the burden placed on the licensee if the requirements

were imposed. This report provides evaluations of the various requests

for relief by the licensee of the Point Beach Units 1 and 2. It deals

only with inservice examinations of components and with system pressure tests.

Inservice tests of pumps and valves (IST programs) are being evaluated separately.

The revision to 10 CFR 50.55a, effective November 1, 1979, modified the

time interval for updating ISI programs and incorporated by reference a later

edition and addenda of Section XI. The updating intervals were extended from

40 months to 120 months in order to be consistent with intervals as defined

in Section XI.

* Hereinafter referred to as Section XI.

For plants with Operating Licenses issued prior to-March 1, 1976, the

provisions of the November 1, 1979 revision are effective after September 1, 1976 at the start of the next one-third of the 120-month interval. During

the one-third of an interval and throughout the remainder of the interval,

inservice examinations shall comply with the latest edition and addenda of Section XI, incorporated by reference in the Regulation, on the date 12 months priQr to the start of that one-third of an interval. For Point Beach Units

1 and 2, the ISI program, and the relief requests evaluated in this report, cover the last 40 months of the current 120-month inspection interval, i.e., from August 1977 to December 1981 for Unit I and June 1979 to September 1982

for Unit 2. These programs were based upon the 1974 Edition of Section XI of the ASME Boiler and Pressure Vessel Code with Addenda through the-.Summer of 1975.

The two reactor facilities, Point Beach Unit 1 and Unit 2, are essentially

identical. The main differences in their ISI programs are that some of the items

in Unit 1 scheduled for examination during the third forty-month period were

scheduled for an earlier period in Unit 2 and vice versa.

The November 1979 revision of the Regulation also provides that ISI programs

may meet the requirements of subsequent code editions and addenda, incorporated by reference in paragraph (b) and subject to Commission approval. Portions of such

editions or addenda may be used provided that all related requirements of the respective editions or addenda are met. These instances are addressed on a caseby-case basis in the body of this report.

References (1) to (22) listed at the end of this report pertain to information transnittals on the.Inservice Inspection (ISI) Reports on Units land 2 between the licensee, Wisconsin Electric Power C=pany, and the Nuclear Regulatory Commission (NRC).

By letters of April 26 and November 22, 1976(1)(3), the NRC provided general ISI

guidance. Technical Specifications changes in response to that guidance were made

by the licensee on February 17, 1977 for Unit 1.(4) ISI program submittals were

made on May 20, 1977(5) for Unit I and on February 26, 1979(13) for Unit 2. The

NRC granted interim relief to Unit 1 on August 26, 1977(7). By letters of August 3,

1977 and December 4, 1978(6)(11) the NRC requested additional information to complete

the report for Unit 1. This information was provided by the licensee on October 6.

*The Point Beach Unit 1 first ten-year interval was extended to December 1981 to

permit inspections to be concurrent with plant outages as allowed in Article IWA2400 of Section XI.

"2

1977 and February 6, 1979. (8)(12) By letters of October 4, 1979 and March 12,

1982(14)(21) the NRC requested additional information on both Units 1 and 2.

This information was provided by the licensee on December 14, 1979 and April 14,

1982.(05)(22) In addition the licensee submitted a 10-year ISI completion

report on Unit I on February 23, 1982.(20)

From these submittals, a total of 7 requests for relief from code require

ments or updating to a later code were identified. These requests are evaluated

in the following sections of this report.

-3-

I. CLASS 1 COMPONENTS

A. Reactor Vessel

1. Nozzle-to-Shell Welds and Nozzle Inside Radiused Sections

(Applies to Unit 1), Category B-D, Item B1.4

Code Requirement

A volumetric analysis of these welds shall be made according to the schedule given in paragraph IWB-2411, which states, "at least 25% of the required examinations shall have been completed by the

expiration of one-third of the inspection interval (with credit for no more than 33-1/3% if additional examinations are completed) and

at least 50% shall have been completed by the expiration of two

thirds of the inspection interval (with credit for no more than

66-2/3%). The remaining required examinations shall be completed by the end of the inspection interval."

Code Relief Requested

Relief is requested from the schedule given in IWB-2411.

Proposed Alternative Examination

All nozzles will be examined once every 10 years when the core

barrel is removed.

Licensee's Basis for Requesting Relief

There are six nozzles in the reactor vessel; two inlet, two outlet and two safety injection. The original intent, as reflected in the technical specifications, was to examine the two outlet nozzles during the first inspection period, the safety injection nozzles during the second period, and the inlet nozzles during the third period. There is no access to the nozzle to vessel welds from the outside-of the reactor vessel.These welds are examined from the inside using a reactor vessel inspection device (PaR Device). Using the PaR Device the core barrel must be removed to provide access to the inlet nozzles. Removal of the core barrel requires a complete unloading of all nuclear fuel from the reactor vessel. This is done only once during each 10-year inspection interval.

The outlet nozzles and nozzle-to-vessel welds were examined from the inside of the nozzles on schedule during the first inspection period. During the second inspection period it became necessary to remove the core barrel in. order to inspect the vessel beltline welds. The safety injection-nozzles were inspected during the second period because of the technical specification requirement. The inlet nozzles were inspected during the second period because the core barrel was removed.

The reactor vessel inspections performed during the second period. were performed in accordance with the 1974 code. The 1974 code increased the requirements for inspection of the nozzle to shell welds from those contained in previous codes. In order to better fulfill the increased requirements of the

code and to provide a better test, the method of performing "-4-

the test was changed from that employed during the first period. A "windmill" device was constructed for use on the PaR Device which enabled inspection of the welds by "scrubbing" the vessel walls in addition to inspecting from the bore of the nozzles. The "windmill" device and this method of inspection are only possible if the core barrel is removed. The outlet nozzles were reexamined using this method during the second period so that all six nozzle-to-vessel welds were examined during the second period. The better test method made possible by performing these examinations with the core barrel removed provides a positive effect on safety.

Evaluation

The schedule for examining welds in the nozzle-to-shell welds and nozzle inside radiused sections for Point Beach Unit I originally was as follows:

1st period (40 months) -- 2 Outlet nozzles

2nd period (40 months) -- 2 Safety nozzles

3rd period (40 months) -- 2 Inlet nozzles

(core barrel must be removed).

During the second period it was necessary to remove the core barrel and all six nozzles were inspected using the "windmill" device. Removing the core barrel and reexamining the two inlet nozzles merely to comply with the schedule clearly is not practical from the standpoint of the personnel exposure incurred for only a marginal gain in safety. The total quantity of welds examined during the interval exceeds the requirements since the outlet nozzles were examined twice.

Conclusions and Recommendations

Based on the above evaluation, it is concluded that for the welds discussed above, the code requirements are impractical. It is further concluded that the alternative examination discussed above will provide necessary added assurance of structural reliability. Therefore, the following is recommended:

Code relief from IWB-2411 should be granted and the proposed alternative of examining all six nozzles at -one time when the core barrel is removed should be approved.

References

Reference 11, pg 1; reference 12, pp 1 and 2.

-5-

2. Safety Injection Nozzle to Safe End Welds (Applies to

Unit No. 1), Category B-F, Item Bl.6

Code Requirement

Volumetric and surface examination covering the cir

cumference of 100% of the welds during each inspection interval. Examinations in each 40 month period shall be in accordance with paragraph IWB-2411.


Request relief from the surface examination and from

the requirements of IWB-2411.


None


This inspection is not practical because these welds

are not accessible from the outside. These welds were

previously examined ultrasonically from the bore.

Evaluation

There are six nozzles in the reactor vessel; two inlet,

two outlet and two safety injection. A surface dye penetrant

examination is performed on all of the reactor nozzle safe

end connections with the exception of the two safety injectionnozzle connections. Neither a visual nor surface examination

can be performed on these welds since they are enclosed by a

concrete sleeve. An ultrasonic examination of the welds of

all six nozzles was performed from the inside diameter of the

nozzles during the second period with the aid of a remote

control examination device (PaR device) along with the nozzle

to shell weld examination.

T'he once every 10-year volumetric examination provides

sufficient information as to the condition of the safety in

jection safe-end weld. Surface examinations are made on the

other four safe-end welds in accordance with the schedule

given in IWB-2411, which does not necessarily match the timing

of the once every 10-year volumetric examination. Should any of

these surface examinations indicate a surface flaw, investigation

should include an additional volumetric examination of not only

the problem weld but also the two safety injection nozzle welds.

If weld cracking should be more general than one specific weld,

this investigation would provide timely information on the

condition of the safety injection welds.

-6-


Based on the above evaluation, it is concluded that for the

welds discussed above, the code requirements are impractical.

It is further concluded that the alternative examination dis

cussed above will provide necessary added assurance of structural

reliability. Therefore, the following is recommended:

Code relief from the surface examination of the safety

injection safe-end welds should be granted. However, if the

surface examination of any of the other reactor safe-end welds

should indicate a surface flaw, an additional volumetric exami

nation of not just the problem weld, but also of the two safety

injection welds should be done during the shutdown.

References

Reference 5; reference 6, pg 3; reference 8, pg 13;

reference 11, pg 1; reference 12, pp 1 and 2.

-7-

3. Integrally Welded Supports (Applies to Unit No. 1),

Category B-H, Item B1.12

Code Requirement

A volumetric examination of a 100% of the welding shall be done according to a schedule given in paragraph IWB-2411.


Code relief is requested from the schedule given in IWB-2411.


The volumetric examination of the reactor vessel welded

supports would be done at once every 10 years when the core barrel is removed.


There are two integrally welded reactor vessel supports which must be inspected. These vessel supports are not accessible from the outside of the reactor vessel. They are inspected from the inside of the vessel using the PaR Device. The core barrel and consequently the nuclear fuel must be removed from the reactor vessel in order to perform these inspections; therefore, it is not practical to split the inspections among different periods. These inspections were originally planned for the third period but were moved to the second period when it became necessary to remove the core barrel ahead of schedule. These tests were performed during refueling outage V#4. It is anticipated that they will be performed during the second peri.cd in succeeding intervals.

Evaluation

The integrally-welded reactor vessel supports cannot be examined from outside the reactor vessel, and must be remotely

examined from inside the reactor using the PaR Device. This

examination can only be performed when the core barrel is removed.

"Fiiis fact makes compliance with IWB-2411 impractical.

Conclusions and Recornendations

Zased on the above evaluation, it is concluded that for the

welds discussed above, the code requirements are impractical. It is further concluded that the alternative examination discussed

above will provide necessary added assurance of structural reliability. Therefore, the following is recommended:

-8-

Code relief from ItWB-2411 should be granted. The alternative

schedule of examining all the integrally-welded reactor vessel sup

ports at least once every ten years should be approved.

References

Reference 11, pg 1; reference 12, pg 3.

4. Vessel Cladding (Applies to Unit 1), Category B-I-I, Item B1.14

Code Requirement

A visual examination of six patches (each 36-s. -in.)

evenly distributed in the accessible section of the vessel

shell shall be made according to paragraph IWB-2411.


Relief is requested from the schedule given in IWB-2411.


The examination on all of the patches would be done at

one time when the core barrel is removed.


Only a small amount of vessel cladding, less than seven

inches, is visible above the core support structure during a

normal refueling. At the time the cladding patches were ex

amined it was felt that examination of locations below the

core support ledge would be more meaningful than examinations

in the normally visible area above the core support structure.

Six cladding patches all below the core support ledge were ex

amined during the second period duiring refueling 4.

Evaluation

To comply with the cladding examination requirements would

require removal of fuel an6 the core barrel. This is an imprac

tical requirement with a relatively small compensation in safety.

The examination that was completed at Point Beach Unit 1 during

the second period is entirely adequate. Examination of nozzle

to-vessel welds covers sufficient cladding in more suspent areas

of the vessel to determine the cladding condition.

The 1977 Edition of Section XI has been referenced in 10

CFR 50.55a and inservice exa-niinations may meet the requirements

-9-

of this edition in lieu of those from previous editions with

the following provisions:

(a) Commission approval is required to update to the more

recent edition (pursuant to 10 CFR 50.55a(g)(4)(iv));

(b) When applying the 1977 Edition, all of the addenda through

Summer 1978 Addenda must be used;

(c) Any requirement of the more recent edition which is

related to the one(s) under consideration must also be met.

The requirements for examining closure-head cladding and

vessel cladding are deleted from the 1977 Edition with addenda

through Surmier 1978.

Recommendations

Pursuant to 10 CFR 50.55a(g)(4)(iv), approval should be

granted to update to the requirements of the Summer 1973 Addenda

for Category B-I-i items. This approval would delete the

requirement to examine these items.

References

Reference 11, pg 1; reference 12, pp 3 and 4.

-10-

B. Pressurizer

No relief requests.

C. Heat Exchangers and Steam Generators

1. Regenerative Heat Exchanger, Integrally Welded Supports

(Units 1 and 2), Category B-H, Item B3.7

Code Requirement

Volumetric examination of 10' of the circumference of the weld to the vessel during each inspection interval.

Code Relief Request.ed

Request relief from making a volumetric (ultrasonic)

examination of the weld.


Visual examination of the accessible portion of the welds.


Ultrasonic examination of the support-to-vessel tack welds

is not practical because of the curvatures of the vessel end

caps. Liquid penetrant examination of these welds is not practical

due to masking caused by penetrant entrapment between the support

member and vessel shell. Radiation levels around the residual heat exchanger are 2R to 3R.

Evaluation

The regenerative heat exchanger welded supports are three

1" long partial-length fillet welds between the heat exchanger

and the saddles. This weld configuration does not lend itself

to ultrasonic examination. Dye penetrant surface examination

is not practical as an alternative examination, since the pene

trant would be trapped in areas adjacent to the weld.

When the developer is applied relatively large amounts of

penetrant would flow out of the adjacent areas and overshadow

any penetrant which might be present as a surface flaw. A

visual examination is the only practical method. The examina

tion period will have to be relatively short since the radiation levels in the area are about 2-3R.

-11-


Cased on the above evaluation, it is concluded that for the welds discussed above, the code requirements are impractical. It is further concluded that the alternative examination discussed above will provide necessary added assurance of structural reliability. Therefore, the following is recommended:

Code relief should be granted from the volumetric examination provided the proposed alternative visual examination is performed.

References

Reference 5; reference 6, pg 4; reference 8, pg 15; reference 13, pg 2.7.

D. Piping Pressure Boundary

No relief requests.

E. Pump Pressure Boundary

1. Reactor Coolant Pumps, Integrally Welded Supports (Applies to

both Units 1 and 2), Category B-K-i, Item B5.4

Code Requirement

A volumetric examination of 25% of 'integrally-welded supports each inspection interval.


Request relief from making a volumetric examination (ultrasonic) of the integrally welded support.


Visual examination.


Volumetric examination of these welds is not practical. The surface is rough and ultrasonic waves do not propagate well in the cast stainless material.

-12-

Evaluation

The ultrasonic examination of the reactor coolant pump

support lugs is not practical because the ultrasonic waves do

not propagate well through the rough cast stainless steel

material. A dye penetrant examination as an alternative in

spection is also not practical because of the rough surface.

After the penetrant has been allowed to dwell, it cannot be

cleaned properly. When the developer is applied to the examina

tion area, the trapped penetrant will be drawn out. Numerous

false indications will appear, and any flaw indications will

be indistinguishable. A visual examination appears to be the

only suitable alternative examination.


Based on the above evaluation, it is concluded that for the

welds discussed above, the code requirements are impractical. It

is further concluded that the alternative examination discussed

above will provide necessary added assurance of structural reli

ability. Therefore, the following is recomended:

Code relief should be granted from the volumetric examina

tion provided the proposed alternative of visual examination is

performed.

References

Reference 5; reference 6, pg 3; reference 8, pg 14;

reference 13, pg 2.11.

-13-

2. Reactor Coolant Pumps, Pump Casing Welds (Applies to both Units

1 and 2), Category B-L-1, Item B5.6

Code Requirement

A volumetric examination of the weld metal and base metal for

one wall thickness beyond the edge of the weld. The examination performed during each inspection interval shall include 100' of

the pressure-retaining weld in at least one pump of each group. The examination may be performed at or near the end of the inspection interval.


Request relief from making an examination oi_one full wall

thickness beyond the edge of the weld on each side.


Examine the weld for a distance of 1/2-inch on either side

of the weld per 1977 Edition, through Summer 1978 Addenda.


For the reactor coolant pump casing weld examination, examining one wall thickness on either side of the weld would

mean that a 17 to 20-inch band would have to be inspected for

each weld. The technique and equipment which has just recently

been developed to perform an inspection of the reactor coolant pump casing welds is not capable of examining bands of this width. Multiple shots using the miniature linear accelerator (MINAC) would have to be taken to cover these band widths. Thus,

instead of about 35 shots per weld, about 100 shots for each of

the three welds would have to be taken. Each shot requires an

exposure time of one-half to three hours. There would be a sub

stantial increase in the accumulated radiation exposure associated

with the placement of the radiographic film and a substantial

increase in the examination time with no additional benefits in

flaw detection. As recognized by the 1977 Edition, an examination

of 1/2-inch of the base metal on either side of the weld encompasses

the expected flaw zone for the reactor coolant pump casing welds.

In this submittal, the licensee also described the results

of the MINAC examination(s) to date and future plans as follows:

In accordance with the inspection items.B5.6 and B5.7, radio

graphic examination of the Unit 1 "B" RCP casing welds and a

visual examination of the pump inside pressure-retaining surface

using the MINAC and manipulator was performed during the Unit 1

1981 refueling outage. Essentially, 100% of all the casing welds

were examined. The only areas not radiographed were the areas

under the pump support lugs and inaccessible portions of the

-14-

discharge nozzle. The MINAC was first utilized at the Ginna Plant.

In MINAC examinations performed at Point Beach, Turkey Point, and

Ginna, no notable indications were found in any of the pumps examined.

The casing examination at Point Beach took about 25 days to

perform, including'the associated pump disassembly and reassembly,

and resulted in a total accumulated radiation exposure of 36 man

rem and a cost on the order of $700,000. Prior to performing the

examination on one of the Unit 2 reactor coolant pumps which are

identical to those of Unit 1, an evaluation of the improvements

in the inspection methods employed will be performed to determine

if the total cost in outage time, exposure, and money can be

reduced to a level more commensurate with the benefits of the

examination. Current plans are to disassemble a Unit 2 RCP and

perform the casing weld examination during the 1983 refueling

outage, or a waiver will be requested after the zepsu]ts from

the H. B. Robinson examination (being performed in April 1982) are available.

Evaluation

The examination required by the 1974 Code is not practical.

No utility has done an inservice inspection of the reactor coolant pump casing welds in accordance with the requirements of the

1974 Edition of ASME Section XI. The diffuser of the pump makes

it impossible to examine one weld thickness above the upper weld

and the changing casing thickness, coupled with the physically allowed beam angles, makes examining one wall thickness on either side of any of the welds difficult.

The 1977 Edition of Section XI has been referenced in

10 CFR 50.55a and inservice examinations may meet the requirez ments of this edition in lieu of those from previous editions with the following provisions:

(a) Commission approval is required to update to the more

recent edition (pursuant to 10 CFR 50.55a(g)(4)(iv));

(b) When applying the 1977 Edition, all of the addenda through Summer 1978 Addenda must be used;

(c) Any requirement of the more recent edition which is related to the one(s) under consideration must also be met.

The MINAC examination of the Unit 1 pump casing weld complied

with the more recent code. A Unit 2 pump casing weld will also be

examined according to the more recent procedure, or else a waiver will be requested, later.

-15-

Recommendations

Pursuant to 10 CFR 50.55a(g)(4)(iv), approval to update to the 1977 Edition (through S-78 Addenda) should be granted. This would require only one-half-inch of base metal on each of the weld to be examined. The examination completed on Unit 1 would then be in compliance, as would the Unit 2 examination now planned for 1983.

References

Reference 5; Reference 13, pg 2.12; Reference 18.

F. Valve Pressure Boundary

(No relief requests).

II. CLASS 2 COMPONENTS


III. CLASS 3 COMPONE•iTS


IV. PRESSURE TESTS


V. GENERAL


-16-

REFERENCES

1. George Lear (NRC) to Sol Burstein (WE), Point Beach Units 1 and 2, April 26, 1976.

2. Sol Burstein (WE) to Bernard C. Rusche (RRC), Point Beach Nuclear

Plant Use of Code Case 1968 for In-Service Inspection, June 2, 1976.

3. George Lear (NRC) to Sol Burstein (WE), November 22, 1976.

4. Sol Burstein (WE) to Bernard C. Rusche (NRC), Inservice Inspection of

Safety Class Components, Technical Specification Change Request I4o. 42,

Point Beach Nuclear Plant Unit 1, February 17, 1977.

5. Sol Burstein (WE) to Edson G. Case (NRC), Docket No. 50-226, Point Beach

Nuclear Plant Unit 1, Inservice Inspection Program for Safety Class

Components, May 20, 1977.

6. George Lear (NRC) to Sol Burstein (WE), August 3, 1977.

7. George Lear (NRC) to Sol Burstein (WE), August 26, 1977.

8. Sol Burstein (WE) to George Lear (NRC), Docket 50-266, Additional Infor

mation Inservice Inspection Program, Point Beach Nuclear Plant Unit 1,

October 6, 1977.

9. A. Schwencer (NRC) to Sol Burstein (WE), July 12, 1978.

10. Sol Burstein (WE) to Harold R. Denton (NRC), Docket No. 50-301, Inservice

Inspection of Safety Class Components Technical Specification Change

Request No. 58, Point Beach Nuclear Plant Unit 2, November 27, 1970.

11. A. Schwencer (NRC) to Sol Burstein (WE), December 4, 1978.

12. Sol Burstein (WE) to Harold R. Denton (NRC), Docket 50-266 Additional

Information Inservice Inpsection Program, Point Beach Nuclear Plant

Unit 1, February 6, 1979.

13. Sol Burstein (WE) to Harold R. Denton (NRC), Docket No. 50-301, Point

Beach Nuclear Plant Unit 2, Inservice Inspection Program for Safety Class

Components, February 26, 1979.

14. A. Schwencer (NRC) to Sol Burstein (WE), October 4, 1979.

15. A. Schwencer (NRC) to Sol Burstein (WE), October 25, 1979.

16. C. W. Fay (WE) to Harold R. Denton (NRC), Docket Nos. 50-266 and 50-302,

Point Beach Nuclear Plant, Units I and 2, Additional Information Inservice

- Inspection Program, December 14, 1979.

17.. C. W. Fay (WE) to Harold R. Denton (NRC), Docket No. 50-266, bRefuli

Inservice Inspection, Point Beach Nuclear Plant Unit 1, October 6, 1

-17-

j ,

18. C. W. Fay (WE) to H. R. Denton (NRC), Docket Nos. 50-266 and 50-301, Inservice Inspection of Safety Class Components Updating of 10-year Plans, Point Beach Nuclear Plant Units I and 2, October 6, 1981.

19. Timothy G. Colburn (NRC) to Sol Burstein (WE), October 21, 1981.

2D. C. W. Fay (WE) to H. R. Denton (NRC), Docket No. 50-266, Completion of First Ten-Year Inservice Inspection Interval, Point Beach Nuclear Plant, Unit 1, February 23, 19U2.

21. Robert A. Clark (NRC) to C. W. Fay (WE), March 12, 1982.

22. C. W. Fay (WE) to H. R. Denton (NRC), Additional Information on Inservice Inspection Program, Reactor Coolant Pump, Point Beach Nuclear Plant, Units I and 2, Apr'il 14, 1982.

-13-

7590-01

UNITED STATES NUCLEAR REGULATORY COMMISSION



NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSE AND NOTICE OF GRANTING OF RELIEF FROM CERTAIN REQUIREMENTS OF

ASME CODE SECTION XI INSERVICE INSPECTION REQUIREMENTS

The U...,S.,,Nuclear Regulatory Commission (the Commission) has

issued Amenodmdpt No. 63 to Facility Operating License No. DPR-24,

and Amendment No. 68 to Facility Operating License No. DPR-27 issued

to Wisconsin Electric Power Company (the licensee), which revised Tech

nical Specifications for operation of Point Beach Nuclear Plant,

Unit Nos. 1 and 2 (the facilities) located in the Town of Two Creeks,

Manitowoc County, Wisconsin. The amendments are effective 20 days from

the date of issuance.

The amendments revise the language of the Technical Specifications

relating to inservice inspection requirements of safety class components

to conform with 10 CFR 50.55a, the Codes and Standards Rule.

The Commission has also granted relief from certain requirements

of the ASME Code, Section XI, "Rules for Inservice Inspection of Nuclear

Power Plan Components" to the Wisconsin Electric Company. The relief

relates to the Inservice Inspection Program for the Point Beach Nuclear

Plant Units 1 and 2. The ASME Code requirements are incorporated by

reference into the Commission's rules and regulations in 10 CFR Part 50.

The relief is effective as of the date of issuance.

DESIGNATED ORIGINAL

Certified By

8210040427 820831 PDR ADOCK 05000266 P PDR

7590-01

-2

The relief relates to certain inservice inspection requirements,

pursuant to 10 CFR Section 50.55a(g)(6)(i) of the Commission's regula

tions, involving volumetric, visual and surface examinations of piping,

component pressure boundary and support structural integrity.

The application for amendments and requests for relief comply with

the standards and requirements of the Atomic Energy Act of 1954, as amended

(the Act), and the Commission's rules and regulations. The Commission

has made appropriate findings as required by the Act and the Commission's

rules and regulations in 10 CFR Chapter I, which are set forth in the

related Safety Evaluation. Prior public notice of these amendments and

requests for relief was not required since they do not involve a signi

ficant hazards consideration.

The Commission has determined that the issuance of the amendment and

granting of relief will not result in any significant environmental impact

and that pursuant to 10 CFR Section §51.5(d)(4) an environmental impact

statement or negative declaration and environmental impact appraisal need

not be prepared in connection with issuance of this action.

For further details with respect to this action, see (1) the licensee's

applications for amendments dated Febrauary 17, 1977 and November 27, 1978,..

(2) the licensee's letters dated May 20, 1977, October 6, 1977, February 6,

1979, February 26, 2979, December 14, 1979, October 6, 1981 and April 14,

1982, (3) Amendment Nos. 63 and 68 to License Nos. DPR-24 and DPR-27,

(4) the Commission's letter to the licensee dated

and (5) the Commission's related Safety Evaluation including the attached

7590-01

-3

Technical Evaluation Report prepared by Science Applications Incorporated.

All of these items are available for public inspection at the Commission's

Public Document Room, 1717 H Street, N.W., Washington, D. C. 20555 and at

the Joseph Mann Library, 1516 16th Street, Two Rivers, Wisconsin 54241.

A copy of items (3), (4) and (5) may be obtained upon request addressed

to the U. S. Nuclear Regulatory Commission, Washington, D. C. 20555,

Attention: Director, Division of Licensing.

Dated at Bethesda, Maryland this 31st day of August, 1982.



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