ML20002C418

Forwards Proposed Change 30 to Tech Specs,Incorporating Facility Inservice Insp Program Per Dj Skovholt 720328 Request

ML20002C418
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/21/1972
From:	Sewell R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8101100277
Download: ML20002C418 (14)

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April 21, 1972

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7 atn p 4 Dr. Peter A. Morris, Director Re:

Docket No 50-155 Division of Reactor Licensing License DPR-6 United States Atomic Energy Proposed Technical Commission Specification Washington, DC 20545 Change No 30

Dear Dr. Morris:

Transmitted herewith are three (3) executed and thirty-seven (37) conformed copies of a request for change to the Technical Spec-ifications of License DPR-6, Docket No 50-155, issued to Consumers Power Company on May 1, 1964 for the Big Rock Point Plant.

This proposed change (No 30) will incorporate the Big Rock Point Plant in-service inspection program into the Technical Spec-ifications as requested in Mr. Donald J. Skovholt's letter of March 28, 1972.

This in-service inspection program has been developed by research of Section XI of the ASME Boiler and Pressure Vessel Code, Y ailed review of piping and equipment fabrier tion drawings, and Mderation of plant physical restraints. Previous in service in-crections have revealed that, in some cases:.omponents have been mislabeled as both inspectable and noninspettable. The program herein submitted is felt to be accurate; however, it is likely that changes wili be necessary as more experience is gained in future rears.

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9 Y p [6f"d RBS/dmb Ralph B. Sewell Z

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Docket No 50-155 Request for Change to the Technical Specifications Change No 30 License No DPR-6 Mr. Donald J. Skovholt's letter of March 28, 1972 requested the following changes to the Technical Specifications of License DPR-6, issued to Consumers Power Company on May 1, 1964, for the Big Rock Point Plant:

A.

Add a new section which will be designated as "Section 9" to read as follows:

'.'9 0 PRIMARY p dTEM SURVEILLANCE 91 APPLICABILITY Applies to in-service structural surveillance of primary system components.

92 ORTECTIVE To insure the integrity of the reactor pressule boundary.

93 SPECIFICATIONS (a) Post-operational inspections shall be made according to the methods and intervals indicated in Table 9 3(a) and IS-242 of Section XI of the AGE Boiler and Pressure Vessel Code, 1971 edition, with summer 1971 addenda.

(b) The structural integrity of the primary syctem boundary shall be maintained at the level required by the ori;;inal acceptance standards throughout the life of the plant.

Any evidence as a result of the tests outlined in Table 9 3(a), that defects have developed or grown shall be investigated, including evaluation of comparable areas of the primary system.

(c) Sufficient records of each inspection shall be kept.to allow comparison and evaluation of future tysts.

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2 (d) A surveillance program to monitor radiation-induced changes in the mechanical and impact properties of the reactor vessel material shall be maintained as required by Section 4.1.1(h).

94 BASIS This inspection program implementsSection XI of the 1971 ASME Boiler and Pressure Vessel Code to the maximum extent practical.

It is recognized that plant design and construction were com-pleted approximately seven years prior to the development of Section XI and it is therefore not possible to comply fully with the code. Areas of exception in the Big Rock Point In-Service Inspection Program are based primarily on experience gained during the three previous in-service inspections con-ducted prior to formulation of this program.

Based on this previous experience, a detailed review of piping and equipment fabrication drawings, plant physical constraints ( } and Sec-tion XI, Table 9 3(a) was developed to summarize the plant in-service inspection requirements. Significant exceptions to Section XI are as follows:

Paragraph IS-211.2 Remote visual examination will be used for visual inspections.

Depending on the physical arrangement of the area to be in-spected, it may or may not be possible to obtain the resolution specified in IG-211.1.

In any circunstance, the best resolution possible shall be obtained for the inspection.

Paragraph IS-232 No preoperational exenination can be made because the plant has been in operation since 1962. It would.also be impractical to attempt an examination of this nature at this time due to high radiation fields and/or difficulties in gaining access to some areas of the reactor coolant pressure boundary.

I September 29, 1971 letter from R. L. Haueter to Dr. Peter A. Morris v

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3 Table IS-261 Item 1.1 - Longitudinal and circumferential welds in the core region are not accessible because the outside of the reactor vessel is closely surrounded by concrete and

. access from the inside is not feasible due to the proximity of the thermal shield.

Item 1.2 - Longitudit 1 and circumferential welds in the reactor vessel shell are not accessible with existing equip-ment. Merid.ional welds in the lower head are also not accessible because of the core support plate and poor geometry due to 40 penetrations.

Item 1 3 - We shall attempt to inspect reactor vessel to flange veld i' rom flange surface. The weld is 22" below the surface of the flange and attenuation may limit the validity of this inspection.

Item 1.4 Primary nozzle to reactor vessel welds are inacces-sible due to configuration of the reactor vessel in-ternals and/or lack of equipment to perform an in-spection of this nature nor does plant design perwxL inspection of these nozzle welds from the reactor vessel exterior.

Item 1.5 This item has been deleted with the exception of the control rod drive "J" welds since plant design and present technology do not permit a volumetric exam-ination of this type of penetration.

Item 1.6 None of the reactor vessel penetrations, except c11-sure head penetrations, is visually accessible because of piant design.

j Item 1.7 A device is being developed to attempt to inspect j

volumetrically the six reactor vessel steam outlet nozzles primary nozzle to safe-end welds.

These welds are not accessible either visually or by sur-face examination. The two 3" vent nozzles in the reactor vessel head are accessible fdr visual and surface inspections.

Previous attempts to inspect 4

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there welds volumetrically were unsuccessful due to the' pef'$ cal configuration.of the nozzles. All other reactor vessel primary nozzle to safe-end welds are inaccessible because plant design will not allow an inspection of this type.

Item 1.12 - This item has been deleted since plant design does not allow access to reactor vessel support pads.

Item 1.15 - The core support plate brackets are not accessible for either direct or remote visual examination due to the physical configuration of the reactor vessel.

Item 3 1 - The steen drum longitudinal welds will not be inspec-ted. (aey are located near the horizontal plane through the center line of the steam drum and are, therefore, inaccessible due to the high radiation field and the inaccessibility of the sides of the steam drum. The circumferential steam drum welds will be inspected -only within 30 of are from the top of the steam drum.

Item 3 2 - The steam drum primary nozzle to reactor vessel welds and inside radius sections will not be inspected from the drum exterior due to the high radiation field that exists beneath the steam drum and the extreme dif-ficulty in removing insulation from the steam drum.

The high radiation field that exists inside the drum precludes inspection from the interior.

Set on nozzles on clean-up system heat exchanges will not be inspected volumetrically due to geometry.

Item 3 3 - The safe-end to nozzle weld on the steam drum vent line will not be inspected volumetrically because its geometry is such that it does not allow proper placement of an ultrasonic probe.

Item 3.6 - The entire inspection of steam drum integrally welded supports will be concentrated on the upper supports i.

due to the extremely high radiation drea under the steam drum.

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5 Item 4.1 - The fo_ lowing safe-end to pipe and c1rbon steel _to

. stainless steel transition welds will not be in-spectedduetotheplantphysicallayoutand/or high radiation areas:

Transition welds:

6" shutdown system outlet line.

4" core spray line.

. Safe-end to pipe veld:

Reactor vessel recirculating inlet nozzles.

Reactor vessel instrument nozzles.

Reactor vessel emergency coolant inlet nozzle.

Reactor vessel unloading outlet nozzle.

Reactor vessel poison inlet nozzle.

Steam drum vent nozzle - will inspect visually and surface, not volumetric (refer to Item 3 3).

Reactor vessel steam outlet nozzles - will inspect volumetrically only.

Item 4.2 - The longitudinal welds on the main recirculatin6 system steam risers are inaccessible due to plant ~

design and high radiation areas.

Some circumferential welds in portions of the var-ious systems that i:omprise the reactor coolant pres-cure boundary are inaccessible due to plant physical designand/orhighradiationfields.

Item 4.3 - Not applicable.

Item 4.5 - Approximately 50% of the integrally velded pipe supports on the main recirculating system are in-accessible due to physical location and high radia-tion areas and will not be inspected.

Item 4.6 - The piping restraints on the downcomers will not be inspected due to physical plant layout.

Item 5.1 - Not applicable.

Item 5 3 - Not applicable.

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' Item 5 5 - Not applicable.

-Item 5.6 - Not applicable.

Item 6.1 - Not applicable..

Item 6.2'

- Main recirculating pump discharge and butterfly

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valves are not isolable from the reactor; therefore, they will not be inspected.

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Table 9 30)

PRIMARY COOLANT SYSTEM SURVEILLANCE-Item No (Corresponds to Examination i

Numbers in Category

.I Table IS-261)

Table IS-251

. Components and W rts To Be Examined Method Reactor Vessel and Closure Head 1.2

-B Closure Head Welds

. Volumetric 13 C

Reactor Vessel to Flange Weld Volumetric Closure Head to Flange Weld Ditto 4

1.5 E-1 Control Rod Drive Penetration "J" Welds

- Volumetric Closure Head Nozzle Welds Ditto 1.6 E-2 Closure Head Nozzle Welds Visual 1.7 F

Primary Nozzle to Safe-End Welds:

Reactor Vessel Steam Outlet Nozzles Volumetric Access Ports in Closure Head

. Volumetric, Visual' and Surface 1.8 G-1 Closure Studs and Nuts Volumetric and'

. Visual or Surface 19 G-1 Ligaments Between

aded Stud Holes Volumetric 1.10 G-1 Closure Washers Visual 1.11 G-2 Pressure Retaining Bolting Visual-4 1

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f Item No (Corresponds to Examination Numbers in Category Table IS-261)

Table IS-251 Components and Parts To Be Examined Method Reactor Vessel and Closure Head (Contd) 1.13 I-1 Closu2e Head Cladding Visual and Surface f,

or Volumetric 1. 1 11 I-1 Reactor Vessel Cladding _

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Visual 1.15 N

. Reactor Ve.ssel Thermal Shield Support Brackets and

. Interior Scrfaces Visually Accessible Visual Heat Exchangers and Steam Drum 31 B

.Longitudinaland/orCircumferentialWeldsfor:

Visual and Volumetric Clean-Up Regenerative Heat Exchangers Ditto Clean-Up Nonregenerative Heat Exchangers-Clean-Up Demineralizer Tank Steam Drum (60 Arc at Top of Circumferential Welds)

Emerency Condenser 32 D

Primary Nozzle to Vessel Head Welds for:

Visual and Volumetric Set-In Nozzles on Clean-Up Regenerative Heat Exchang Ditto Clean-Up Demineralizer Emergency Condenser Steam Outlet and Relief Valve Nozzles on Steam Drum 33 F

Primary Nozzle to Safe-End Welds Visual and Surface and Volumetric

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3.4 G-1 Pressure Retainirg Bolting Visual and Volumetric

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Item No (Corresponds to Examination Numbers in Category Table IS-261)

Table IS-251 Components and Parts To Be Examined Method Heat Exchangers and Steam Drmn (Contd)

-35 G-2 Pressure Retaining Bolting Visual 3.6 H

Integrally Welded Vessel Supports Near Top of Steam Drum

' Visual and Volumetric 37 I-2 Steam Drum Cladding Visual Piping Pressure Boundary t

14. 1 F

Safe-Ends or Dissimilar Metal Welds:

Transitions at Clean-Up System Heat Exchangers Visual and Surface and Volumetric Transition at Shutdown System Inlet Ditto Transition at HCV-30 in Control Rod Drive System Control Rod Drive Sy.atem Inlet to Clean-Up System Bypass on Control Rod Drive Pump to Coding Header Bypass Reactor Vessel Steam Outlet Nozzles Volumetric Steam Drum Downcomer Nozzles Visual and Surface and Volumetric Steam Drum Riser Nozzles Ditto Steam Drum Vent Nozzles Visual and Surface o

Item No (Corresponds to Examination Numbers in Category Table IS-261)

Table IS-251 Components and Parts To Be Examined' Method Piping Prensure Boundary (Contd) 3 4.2 J

Cire mferential Pipe Welds:

Visual and Volumetric ~

C No of Welds To Be Inspected Estimated No of During Inspec-Accessible System tion Interval

' Welds Main Steam 13 32 Feed-Water 14 45 Emergency Condenser 13 45 Clean-Up 40 122 Shutdown 11 34 Main Recirculating 22 68 Core Spray 4

14 Redundant Core Spray 9

33 Poison >2" 12 22 4.4 G-2 Pressure Retaining Bolting <2" Vit.ual 4.5 K-1 Integrally Welded Pipe Supports (Except Those Above Visual and Volumetric

-585' Level in Recirculating Pump Room)

, 4.6 K-2 Piping Support and Hanger (Fixcept Those Between 595' Level and 636' Level in Recirculating Pump Room)

Visual 4.7 J-2 Circumferential and Longitudinal Pipe Welds and Brreich Pipe Connection Welds Visual 4.8 J-1 Socket Welds 2" and Greater Visual and Surface go

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Numbers in' Category

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Table IS-251 Componente and Parts To Be Examined Method Pump Pressure Boundary 5.2 L-2 Pump Casings Visual p

5.5 G-2 Pressure Retaining Bolting <2" Visual 57 K-2 Supports and Hangers Visual-Valve Presctre Boundary 6.2 M-2 Valve Bodies

. Visual 6.3 F

Valve-to-Safe-End Welds Visual and Volumetric-6. 15 G-1 Pressure Retaining Boli,ing >2" Visual and Volumetric 6.5 G-2 Pressure Retac11ng Bolting <2" Visual 1

6.6 K-1 Integrally Welded Supports Visual'and Volumetric 6.7 K-2 Supports and Hangers Visual"

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12 B.

Change Section 1.1.2 to read as follows:

"1.1.2 Sections 1 through 7 present the plant operating limits (in-cluding original fuel) through the period of the full-term license. In the order of presentation, Section 1.0 concerns the introduction, Section 2.0 concerns the site, Sections 3 0 through 6.0 cover the plant and its systems, and Section 7 0 presents procedures for plant start-up and procedures for nor-mal and emergency operation of the plant. This section also includes administrative and procedural safeguards to the extent that these have a potential effect on nuclear safety. Section 8.0 describes Phase II of the Research and Development Program, including parametric variations and operating limits on R&D fuel during Phase II.

Section 9 0 presents the primary system surveillance requirements."

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Add the following at the end of the Table of Contents:

"9 0 Primary System Surveillance.

81" CONSUbERS POWER COMPANY By Senior Vice President Date: April 21, 1972 Sworn ana subscribed to before me this 21st day of April 1972.

h_M e G.Unch ]

Notary Public, Jackson County, Michigan btr Commission Expires December 8, ISr[5 i

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Eequest for change to the Technical Spces, Change No. 30, notarized 4-21-72.

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