ML20072C443

Amend 1 to Steam Generator Repair Rept

ML20072C443
Person / Time
Site:	Point Beach
Issue date:	03/01/1983
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20072C431	List: ML20072C429 ML20072C443
References
NUDOCS 8303080435
Download: ML20072C443 (89)

Text

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POINT BEACH STEAM GENERATOR REPAIR REPORT O AMENDMENT 1 INSTRUCTION SHEET The following instructional information and check list is being furnished O to help insert Amendment 1 into the Point Beach Steam Generator Repair Report.

Discard the old sheets and insert the new sheets, as listed below. Keep O '

these instruction sheets in the front of the report to serve as a record of change. ,

, REMOVE INSERT i (Front /Back) (IYont/Back) 1/-- 1/--

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TABLE OF CONTENTS OV Page No.

1.0 INTRODUCTION

, SIM4ARY AND CONCLUSIONS i- 1.1 Summary of Steam Generator Repair Program 1-1 1.1.0 Introduction 1-1 1.1.1 Containment Entry and Exit of Steam i

Generator Lower Assemblies 1-2 1.1.2 Steam Generator Lower Assembly Characteristics 1-2

, 1.1.3 Safety Related Considerations 1-2 l.1.4 ALARA Considerations 1-3 1.1.6 OfCite Radiological Considerations 1-3 1.1.6 Other Aspectr. of the Program 1-3 4

1.1.7 Steara Generatar Disposal 1-4 1.2 Identification of Prine.ipel Agents and Contractors 1-4 ,

l.3 Other Cce:ideratitas 1-4 f I 1-5 1.4 Canclusions 2.0 REPLACEENT C0FF0ENT DESIGN 2.1 Comparison with Existing Component 2-1 2.1.1 Parametric Comparison 2-1 2.1.2 Physical Compatibility with Original Steam Generators and Systems 2-2 2.1.3 ASE Code Applications 2-2 i 2.1.4 Regulatory Guide Application 2-2 4

2.2 Component Design Improvements 2-7

2.2.1 Design Refinements to Minimize the i Potential for Corrosion 2-7 2.2.2 Design Refinements to Improve Performance 2-9 2.2.3 Design Features to Permit Ease of l

Maintenance and Reliability 2-10A 2.3 Shop Tests and Inspections 2-12 2.4 Quality Assurance 2-12

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TABLE OF CONTENTS (Continued)

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3.0 COMP 0NENT REPLACEENT PROGRAM AND PROCEDURES 3-1 3.1 Pathways and Construction Restrictions 3-2 Site Preparation 3.1.1 3-2 3.1.2 Containment Preparation 3-4 3.1.3 Transportation On-Site 3-5 3.1.4 Rigging Configuration 3-6

] 3.1.5 Rigging and Handling Controls 3-7 l 3.2 Equipment and Concrete Pemoval ar:d Replacement 3-7 i

3.2.1 Mechanical Equipment 3-8 3.2.2 Instrumentation 3-8 3.2.3 Cable and Condtit 3-8 i 3.2.4 Piping 3-9 3.2.5 Concrete and Structural Steel 3-9 3.2.6 Removal and Installatir.n of Steam i Generators 3-10 3.3 Radiological Protection Program 3-20

( 3.3.1 Supplemental Access Control 3-21  !

3.3.2 Laundry 3-23 3.3.3 Control of Airborne Radioactivity and Surface Contamination 3-23 3.3.4 Supplemental Personnel Monitoring  :

l l Requirements 3-24 i I 3.3.5 General ALARA Considerations 3-26  ;

i 3.3.6 Miscellaneous Waste Disposal 3-28

\_) 3.4 Disposition of Steam Generator Lower Assemblies 3-30  !

3.4.1 Objectives of Handling / Disposal Operations 3-30 f

i 3.4.2 Onsite Storage 3-30A l 3.4.3 Offsite Disposal 3-31 l O 3.4.4 Radioactive Releases and Dose Assessment  !

j Associated with Onsite Storage 3-31A {

! 3.4.5 Accident Considerations Associated with l

! Onsite Storage 3-328  !

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TABLE OF CONTENTS (Continued) l Page No.

4 j 7.0 Environmental Aspects of the Repair i 7.1 General 7-1 '

! 7.2 Resources Committed 7-1 ,

7.2.1 Non-Recyclable Building Materials 7-1  !

l 7.2.2 Land Resources 7-1 I

7.2.3 Water Resources 7-3

! 7.3 Waste Water 7-4 7.3.1 Sanitary Facilities 7-4 i l 7.3.2 Laundering Operations 7-5

! 7.4 Corstruction 7-5 ,

I 7.4.1 Ncrse 7-5 7.4.2 Dust 7-5 l 7.4.3 Open Burning 7-6 '

! 7.5 Radiological Aspects 7-6

! 7.6 Return of Operation 7-6 7.6.1 Water Use 7-6 j 7.6.2 Operational Exposure 7-7 l 7.6.3 Radiological Releases 7-7 l 8.0 NRC Questions and Answers 1

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'T LIST OF FI6 ORES Figure No. Title Figure 2-1 Steam Generator Lower Assembly Figure 2-2 Flow Distribution Baffle and Blowdown Figure 2-3 Quatrefoil Tube Support Plate Schenatic Figure 2-4 Tube to Tubesheet Juncture Figure 3-1 Outage Sequence Figure 3-2 Lower Assembly Removal Sequence (Sheet 1 of 2) l Figure 3-2 Lower Assembly Removal Lequence (Sheet 2 of 2)

Figure 3-3 Steam Generator Laydown Areas Figure 3-4 Steam Generator Haul Routes I i

{ Figure 3-5 Reactor Coolant Piping Cut Points

! Figure 3-6 Feedwater and Steam Line Piping Cut Points Figure 6-1 Health Physics Organization Chart O Figure 6-2 PDNP-SGRP HP Interface Chart i

Figure 7-1 Proposed Plot Plan O .

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2. Support plate material has been changed to SA-240 Type 405 from e SA-285 Grade C to minimize corrosion and the potential for denting; and 4

3. The steam generator tube material for the replacement steam genera-tor assemblies is thennally-treated Inconel 600. The original tube Os material was mill-annealed Inconel-600.

Material changes due to design improvements will not degrade the physi-cal, mechanical and thermal performance of the steam generators. Fur-C ther discussion of material changes is provided in Section 2.2 Table 2-2 provides comparison of past and present aplications of materials.

l 2.1.2 PHYSICAL COMPATIBILITY WITH ORIGINAL STEAM 6ENERATORS AND SYSTEMS The replacement steam generator lower assemblies are designed to be duplicate physical mplacements for the existing units. Outside overall dimensions are the same as are the locations of nozzles and support -

attachments. Existing interfaces between the steam generators and plant O

v components and systems are maintained. Dry and wet weights and center of gravity of the steam generators will remain essentially the same; therefore, no changes to the existing supports are necessary.

4 2.1.3 ASME CODE APPLICATION The original steam generators were bLilt to the 1965 Edition of the ASME

Boiler and Pmssure Vessel Code ( ASME Code), including Addenda through

)p Summer 1966; the replacement steam generator lower assemblies vill be

'V designed and fabricated to the latest edition of the ASME Code in effect as of December 1,1979. The stress analysis will be performed using the j 1965 Edition of the ASME Code, including all Addenda through Summer 1966.

2.1. 4 REGULATORY GUIDE APPLICATION i

The compilation below addresses Regulatory Guides considered applicable

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to the fabrication of the mplacement lower assemblies by Westinghouse v Electric Corp. It must be noted that these guides were issued subsequent to construction and 2-2

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1. Westinghouse controls its suppliers to. ,

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b] a. Limit the use of code cases to those listed in Regulatory Position C.1 of the applicable guide revision in effect at the time the equipment is ordered, except as allowed in item 2 below,

b. Identify and request permission for use of any code cases not listed in Regulatory Position C.1 of the applicable guide revision in effect at the time the equipment is ordered, where use of such cases is needed by the supplier.

c. Allow continued use of a code case considered acceptable at the time of equipment order, where such code case was subsequently annulled or amended.

2. Westinghouse seeks WRC pennission for the use of code cases needed by suppliers and not yet endorsed in Regulatory Posi-tion C.1 of the applicable guide revision in effect at the time the equipment is ordered and pennits supplier use only if NRC pennission is obtained or is otherwise assurd (e.g., a later version of the regulatory guide includes endorsement).

1.88 Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records.

1.92 Combination of Modes and Spatial Components in Seismic Response Analysis (Rev.1, Feb.1976).

l.116 QA Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems (May 1977).

1.121 Bases for Plugging Degraded PWR Steam Generator Tubes (April 1977).

1.123 Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants (Revision 1, July 1977).

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l.144 Auditing of Quality Assurance Programs for Nuclear Power Plants.

O 1.146 Qualifications of Quality Assurance Program Audit Personnel for Nuclear Power Plants.

2.2 COMPONENT DESIGN IMPROVEENTS O

The physical, thennal and hydraulic characteristics of the steam genera-tors will be at least equivalent to those of the original steam genera-tors. Additional design features have been incorporated in the design.  !

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2. 2.1. 3 TUBE EXPANSION IN TUBESHEET Following insertion into the tubesheet hole, tack rolling, welding and gas leak testing, the tubes are hydraulically expanded to the full depth of the tubesheet holes. Field experience with operating units in which the tubes were partially rolled into the tubesheet indicated that a full depth O expansion, which would essentially close the crevice between the tube and tubesheet, would add margin by minimizing the possibility of crevice corrosion.

A change in configuration was made in the early 1970's to full depth mechanical rolling which sealed the tubesheet crevice up to approximately a quarter 1.ich from the top of the tube sheet. Full depth mechanically ,

rolleo steam generators now in operation include OHI 1 and KORI 1; these have operated for approximately five years. In that same tinx! period, a number of steam generators with partially expanded tubes were full death expanded ir, the field by an explosive process callea WEXYFX. These include Trojan, Beaver Valley 1, Salem 1, Farley 1, and North Anna 1 and

2. Significant o;:erating experience with full depth expansion is provided I

s by these steam generators, operating for as long as seven years in the case of Trojan.

In the mid-1970's, development effort started on an alternate expansion process that would combine the reduced defonnation and the low residual stress transition of the WEXTEX process with the tight sealing of the hard mechanical roll.

Hydraulic expansion was adopted as the optimum and is the reference Os. process for the current steam generator design including the replacement lower assemblies for the Point Beach Nuclear Plant. The replacement Surry 2 steam generators are the first operating units with hydraulically i expanded tubes. Refinement of the hydraulic process has resulted in s expanding all but a small crevice of about one eighth inch average depth at the top surface of the tubesheet, see Figure 2-4.

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The benefits of the hydraulic expansion process are the reduction of the cold working caused by the mechanical hard rolling and the lower residual V stresses at the transition of the expanded to unexpanded region of the  ;

tubes. Analyses and experiments have shown these tensile stresses to be of the order of 20 ksi on the OD surface and 20-30 ksi on the ID, which are about half the stresses for a mechanical roll.

In addition to the change to hydraulic expansion, the tubing material was also changed to take advantage of the increased corrosion resistance of p thermally treated Inconel 600 to stress corrosion cracking in both primary V and secondary environments. The occurrence of SCC and IGA, which has been observed in some partially expanded units is expected to be minimized oy the combination of the full depth hydraulic expansion and the themal treatment of the Ir.conel 600 tubing. Confimation has been obtained frcm a number of laboratory tests; the results of several tests are summarized as follows.

1. To confim the absence of chemical concentration in the crevice n

v remaining from hydraulically expanding the tube into the tubesheet, a chemical hideout test. was perfomed using a 10 ppm Naps 04 solution (as S0j) as make-up to the bulk secondary fluid of a model ,

boiler. With this concentration, a concentration factor of approximately 20,000 is required for precipitation to occur. Testing of a simulated hydraulically expanded tubesheet joint showed no indication of hideout during testing nor was there any precipitate in the tube examination following testing. These observations are contrasted to the results of similar testing on a partially rolled crevice configuration in which a precipitate on the tube surface in

~-- the vicinity of the top of the tubesheet was present in the post-test examination, indicating a significant concentration within the tube-tubesheet crevice.

Longer term testing of hydraulically expanded tubes in model boilers continues to confirm that the seal between the tube and tube-sheet is adequate and that the upper crevice shows no corrosion. The crevice x dimensions in these tests range from 0.102 in. to 0.179 in, depth.

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2. The effect of the residual stresses on the OD surface of hydraulically expanded tubes in the transition region was assessed in a series of p tests in two aggressive caustic environments. Thermally treated V

Inconel 600 tubes were hydraulically expanded into simulated carbon steel and Inconel 600 tubesheets, internally pressurized to 30,000 psi hoop stress (which is well above the operating stress) and exposed on

(/ the OD to 10% caustic at both 600*F and 650*F. The behavior of the expanded samples was compared to that of thermally treated tubes which were unexpanded.

In the 600*F test solution, none of the samples, expanded or unexpanded, showed cracking after about one year in test. This test temperature is approximately that of the normal hot leg operating l

temperature at Point Beach. At the much more aggressive test temperature of 650*F, which is intended to accelerate the corrosion )

mechanism and provide for more definitive differentiation; unexpanden tube specimens showed cracking in as short as 60 days exposure, ,

whereas the minumum time to leak for hydraulically expanded tubes wa::

261 days. These tests confirm that hydraulic expansion does not p degrade the inherent corrosion resistance of thermally treated Inconel 600 tubes in aggressive caustic enviroments.

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In conclusion, the use of full-depth hydraulic expansion essentially eliminates the tube sheet crevice in which concentration of impuritiet has occurred in the original steam generators and results in an expansion transition which retains the enhanced corrosion resistance of thermally treated Inconel 600 material.

%) 2.2.1.4 THERMALLY TREATED INCONEL 600 TUBING Research by Westinghouse has determined ~ that additional resistance in the l

stress corrosion of Inconel 600 tubing can be achieved by modification of

(.,/ the metallurgical structure through thermal treatment. The primary objec-tive of this treatment is to develop a metallurgical structure, associated 1

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with grain boundary precipitate morphology, which provides increased mar-gin with respect to stress corrosion resistance. Several benefits result from this treatment such as additional resistance to stress corrosion cracking in Na0H, additional resistance to intergranular attack in oxygen-ated environments, additional resistance to intergranular attack in sulphur-containing species and reduction of residual stress imparted by tube processing.

2.2.1.5 0FFSET FEEDWATER DISTRIBUTION Feedwater distribution within the steam generators is modified so that approximately 80 percent of the flow is directed to the hot leg side of the bundle and the remaining 20 percent of the flow is directed to the cold leg side of the bundle. This reduces the steam quality in the hot leg side of the bundle and raises the steam quality in the cold leg side of the bundle. The effect of these changes in steam quality is to shift the point of highest steam quality at the tubesheet elevation toward the cente of the bundle. This area is utilized for location of the blowdown intake. Feedwater flow distribution is accomplished by providing a grea-ter number of flow paths on the portion of the feedwater ring which i

traverses the hot leg side of the tube bundle.

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.2.2.2.1 FLUSH TUBE TO TUBESHEET ELD The tubes on the replacement lower assemblies will be flush with the tube-j sheet-holes and then welded to the tubesheet cladding. Elimination of the l

protruding tube stub of the original design results in lower entry pres-sure loss'es and, therefore, a lower pressure drop in the primary loop. In addition, a possible point of crud buildup and corrosion is minimized with this design. This illustrated in Figure 2-4.

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2.2.2.2 TUBE LAE BLOCKING DEVICE-

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A portion of the recirculated water exiting at the bottom of the wrapper j will tend to preferentially channel to the tube lane and bypass part of j the tube array. In order to minimize this tube bundle bypass, a series of i

j plates are installed in the tut,e lants to block the bypass flow paths.

These plates are compatible with sludge lancing.

l 2.2.2.3 MOISTURE SEPARATOR MODIFICATIONS The secondary moisture separator external drains will be changed to-larger t internal drains. The existing primary separator swirl vane barrels will

! be replaced with a primary moisture separatc,r assembly consisting of one

! hundred and twelve modular 7" I.D. swirl vane assemblies. These modifica-l tions provide improved steam-water separation and reduced moisture

carryover.

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2.2.2.4 FEEDRING MODIFICATIONS Steam nenerator waterhansner is believed to be caused by the rapid condensation of steam within the feedwater line. The presence of steam in the feedwater line can occur during hot standby operation when the steam generator water level drops below the feedring elevation, allowing the feedring to drain and steam to enter the piping. The subsequent addition of cold auxiliary feedwater can cause the steam to condense rapidly, resulting in waterhammer.

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- The Model 44F replacement stea,n generators utilize a feedring designed to limit drainage during such operations by the use of top discharge J-nozzles and a welded feedwater nozzle thermal liner. This configuration

(] has been demonstrated by field experience to effectively limit the poten-tial for water hammer; examples include steam generators backfit with J-nozzles, operating experience with the Model SlF steam generators at j Surry 1 and 2, and operating experience with the Model 44F steam genera-tors at Turkey Point 3 and 4. In addition, the feedwater piping configur-ation at Point Beach Nuclear Plant includes feedwater check valves located close to the steam generator feedwater nozzles. This configuration mini-mizes the potential for significant quantities of steam in the feedwater piping and thus minimizes the potential for waterhammer to occur.

2.2.3 DESIGN FEATURES TO PERMIT EASE OF MAINTENANCE AND RELIABILITY Operational experience, including necessary maintenance and repair, hn led to certain changes in design with the objectives of itureasing reliability and providing additional maintainability of the units. These changes are discussed below and do not affect performance or FSAR safety analyses.

t 2.2.3.1 ACCESS PORTS The replacement lower assemblies are provided with additional access ports. Four 6-inch access ports will be located slightly above the tube-l ba O

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2.2.3.6 STEAM N0ZZLE FLOW LIMITING DEVICE

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A flow limiting device will be provided to be installed in the steam out-let nozzle to minimize the pressure drop across internal components during a postu. ited steam line break transient and also to help minimize the blowdown rate for the postulated accident condition.

2.3 SHOP TESTS AND INSPECTIONS The tests and inspections required by the ASE Code,Section III will be O

Q conducted during the fabrication of the steam generator lower assembly.

In addition to these ASE requirements, further tests and inspections will

' be conducted at the fabrication facility. After the tubing installation is completed a gas leak test will be performed to demonstrate the inte.-

grity of the tube-to-tubesheet welds. The primary side of the steam gen-erator will be hydrotested at the shop in accordance with approved procedures.

2.4 QUALITY ASSURANCE 2.4.1 WESTINGHOUSE WATER REACTOR DIVISIONS (WRD) QUALITY ASSURANCE PROGRAM The design of the replacement steam generators is in accordance with the quality assurance program described in WCAP-8370, Rev. 9A, Amendment 1,

" Westinghouse Water Reactor Divisions Quality Assurance Program" 2.4.2 WESTINGHOUSE NUCLEAR COMPONENTS DIVISION QUALITY ASSURANCE PROGRAM The quality assurance program for the fabrication of the replacement steam generators is in accordance with " Westinghouse Nuclear Components Division Quality Assurance Program Manual," Rev. 5, dated 4/27/82 and WCAP-8370, Rev. 9A, Amendment 1, " Westinghouse Water Reactor Divisions Quality Assurance Program".

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Figure 2-4 Tube to Tubesheet Expansion

3.1.2.4 CONTAINENT STRUCTURAL ANALYSES Containment structural analyses have been performed in accordance with the design criteria in Section 5 of the FSAR for the following:

1. Temporary laydown areas at elevation +66'-0" on the operating floor. These areas will be required to support the upper assem-blies, pipe sections, and miscellaneous construction equipment.

(See Figure 3-3).

2. Containment base mat and existing floor support embeds in the con-tainment wall. Temporary transfer rails will be installed inside the containment to facilitate the removal of the lower assemblies.

Rail loads inside the containment will be transmitted to the base mat. (See Figure 3-2).

These analyses indicate that the containment, foundation and internal structures are capable of supporting the construction loads without permanent modifications to the existing structures.

3.1.3 TRANSPORTATION ON-SITE Movement of the new steam generator lower assemblies (220 tons) on site can be accomplished by several methods such as flatdeck trailer or crawler transporter. Motive power may be rubber-tired tractor or tracked vehicle.

3.1.4 RIGGING CONFICJRATION 3.1.4.1 Inside Containment The upper assemblies will be parted from the lower assemblies and lifted

) and inverted by pad eyes and commercial sling assemblies and relocated to selected storage locations, as discussed in Subsection 3.1.2.3.

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The lower assemblies will be lifted from their compartments using con-ventional hoisting techniques. The hoist lower load block will be linked by pins to a steam generator lift beam equipped with toggle anns or endless grommet type cables. The toggle arn'r, or cables will engage existing lifting trunnions on the assemblies. Each lower assembly will be lifted and transferred in turn to a point approximately 11' from the containment inside wall and approximately on the centerline of the equipment hatch. Special tilting upending /downending skids assemblies, such as Hillman roller units and structural members, will be used to move the assembly from the vertical to the horizontal position. Trans-( fer of the lower assemblies through the equipment hatch will require the connection of additional roller assemblies as each lower assembly travels beyond the reach of the polar crane hoist.

TPe existing overhead trolley is not suitable for steam generator lower assembly replacement, will be moved aside, will not be used for special heavy lifts during the steam generator repair, and will not be modified in any manner. Thus, a load test for the existing overhead trolley will not be r quired. A temporary 250 ton construction hoist will be placed on the polar crane bridge. The temporary hoist will be load tested to meet current OSHA Safety Standards prior to its use for construction lifts. The stresses for the polar crane girders will be maintained within design by utilizing a center post at EL.66'-0". Presently, no structural modifications to the bridge are anticipated. Therefore, a load test for the bridge will not be required. In the event that bridge modifications are required, the lifts performed during the repair pro-ject will constitute an adequate lead test of the modified bridge in l lieu of the load test required by ANSI B30.2.0-1976, paragraph 2-2.2.2.

Special heavy lifts during steam generator repair, as defined by ANSI B30.2.0-1976, will meet the requirements of Section 2-3.2.1 which details the requirements for the handling of special he5vy lifts. Addi-

, tionally, normal industrial and construction standards for the handling i of high value equipment will be observed. As stated in Section 3.2.6.2, the handling of heavy loads in the contairment has no safety impact on plant operation, since all fuel will be removed from the reactor during l

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I the steam generator repair. Therefore, it is considered that the l detailed requirements and procedures which have been implemented at i

Point Beach in response to NUREG-0612 are not applicable to the steam generator repair activities.

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i Reccrds of special heavy lifts will be maintained in prcject records as part of the Control Work Packages discussed in Section 3.2.6.3 and Section 6, paragraph 6.4.

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The inspections indicated in ANSI B30.2.0-1976, Chapter 2-2, will be performed after completion of the steam generator repair and prior to

-l returning the crane to its normal service.

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j 3.1.4.2 OUTSIDE CONTAINENT The lower assembly will exit the containment approximately at grode, on the access area previously described in Section 3.1.1.2. Transfer to a

trailer / crawler transport system will be accomplished by a suitable lifting device.

O Shipping saddles and tie downs will be provided for secure attachment while the transport device is in transit to and from the storage area.

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3.2.4 PIPING In order to accomplish the steam generator repair it will be necessary to cut portions of the following major piping systems: -

A. Reactor Coolant piping.

B. Main steam piping, including small pipe vent lines.

C. . Main feedwater piping.

D. Steam generator blowdown piping.

i The open ends of cut piping will be shielded and as appropriate, covered to ensure cleanliness during the repair.

Reactor Coolant pipe and fittings that are replaced during the Steam i Generator change over will be of type 304 centrifugally cast stainless steel or with the existing plant type 316 pipe material.

Piping weld end preps, welding and nondestructive examination for the reinstallation will be in accordance with the 1977 edition of the ASME i

Boiler and Pressure Vessel Code and Addenda through Summer 1979. The piping system will be reinstalled in accordance with FSAR criteria.

3.2.5 CONCRETE AND STRUCTURAL STEEL The following structures or portions of structures within the contain-ment will be removed to provide a path for the lower assembly:

A. The removable shield wall panels of steam generator cubicles above elevation of 66'-0".

B. A portion of the floor framing and grating at elevation +66'-0" above the equipment hatch.

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i C. A portion of the floor and removable floor slabs at elevation 21'-0"  ;

and 26'-0".

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The upper portion of the steel stairway near the equipment hatch

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1 F tion of the repair outage. The removal of the fuel assemblies will eliminate the possibility of potential incidents involving the fuel which could affect the health and safety of the workers or the general public.

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2. Access to the containment will be through the present. equipment and j personnel hatch; therefore, no structural changes will be mquired I

to the containment structure which forms part of the containment j pmssure boundary. Minor structural changes, e.g., chipping of concrete, may be required on internal walls; however, the effect on 4

internal structures is expected to be insignificant.

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3. The entire repair process will be preplanned. The guidelines con-tained in Hegulatory Guide 8.8, "Infomation Relevant to Ensuring j That Occupational Radiation Exposure at Nuclear Power Stations Will

- Be As Low As Heasonably Achievable", will be followed where appli-cable. In keeping with these guidelines mockups and training will I be used to minimize outage time and radiation exposure. Decon-tamination and other exposure limiting techniques will be used where j

they offer a significant savings in exposure commensurate with over-4 all program objectives. Decontamination consists of removal of i

radioactive contamination, to the extent practical, from work areas

l. of the containment frequented by personnel during the steam genera-1 l tor mpair. The decontamination of these areas will be accomplished

!' using techniques the same as, or similar to, those in use during l nomal plant operation and maintenance. Such techniques include flushing of surfaces with water, mopping, wiping with absorbent j cloth or covering of floor surfaces. Special scaffolding and other

components will be prefabricated to the extent possible to minimize l' radiation exposure and outage time.

p 4. The mactor cavity will be covered by structural members to minimize the possibility of impacting the reactor vessel and associated com-ponents during the repair program. ,

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3-11 e _ _ -- _ _ _ -_ _ _ - - - . - .

5. Procedures will be implemented to preclude the introduction of foreign objects into the steam generators during the repair. These procedures include a combination of physical barriers and admini-i strative controls.' Physical barriers will be specified as part of the Control Work Packages, consistent with the work to be per-i formed. Physical barriers such as herculite, metal plates, and

} decking will be used in work areas, as appropriate.

l.

Administrative procedures will include procedures for personnel access control, tool control and log-in and log-out procedures. The

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administrative controls will include such requirements as lanyards on small equipment and personal items, and design features such as l lock wires on equipment to prevent loss of material in the steam

generator.

6. The repair program will be completed in accordance with the Point Beach Quality Assurance Manual and Section XI of the ASE Code, l

including such items as interaction of repair activities with the 4

unaffected part of the plant, design reviews, radiation control procedures, document control, material acquisitions, etc.

7. Westinghouse will comply with the requirements of the ASE Code Section III, Class I, NB-4620,1977 Edition through the Surener 79 Addenda as applicable for stress relief of each type of weld.

Stress relief heat treatment industry practice, typical of that used at the Turkey Point and Surry plants during steam generator repairs,

-will be used for the Point Beach Unit 1 steam generator repair.

! Special work packages for these activities will be used during the

steam generator repair.

1 i

l The Point Beach steam generator repair methodology does not involve

! any field welded joints between cladded components. The steam generator channel head cladding is applied at the manufacturing

) facility. Safe ends are installed on the inlet and outlet reactor j coolant nozzles at the manufacturing facility to facilitate field i'

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welding of the nozzles to the stainless steel pipe fittings without the need for stress relief heat treatment. Therefore, there will be no welding, cutting, or stress relief heat treatment effects of cladded components.

8. The actual repair process will be similar to the methods used auring V original construction of the units. Much of the experience gained during original construction is applicable to the repair process and will be used as appropriate.

O V 9. The potential environmental effects of the repair program are expected to be minimal. However, reasonable precautions will be exercised to further minimize any environmental impact.

l

10. Presently installed station facilities will be augmented as required to accomodate the additional personnel who will participate in the repair program or to facilitate the actual repair work. The areas of special concern are facilities to prevent the spread of radio-active contamination, disposal of radioactive material, and security

) provisions.

11. The major portion of the repair program will be performed by a com-mercial installer under the direction of Westinghouse personnel.

The installer will provide quality control personnel and procedures and Westinghouse will provide quality assurance personnel. The installer will be required to have an ASME certification as appli-cable to the work he is to perfom.

10. The length of the steam generator repair outage is now estimated to

be approximately 180 days. This schedule is predicated on perform-ing in containment work on two shift rotation that will pennit work-i ing 20 hrs / day. The schedule is divided into the following phases:

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a. Preshutdown activities

b. Shutdown and preparatory activities

c. Removal activities

d. Installation activities

e. Post installation activities

f. Startup activities t g. Post startup activities Each of these phases is discussed in the following paragraphs.

) 3.2.6.3 PRESHUTDOWN ACTIVITIES Prior to the first unit shutdown, the repair program will be preplanned I

and appropriate provisions made for accomplishing each activity required in the repair process. Appropriate procedures, drawings, and instruc-tion will be utilized in the perfomance of repair activities. Engi-neering activities will be completed during this time, as well as estab-lishing temporary installation facilities, material acquisition, train-ing of personnel, prefabrication of certain components and completion of items which do not require a unit shutdown.

The " work package" concept will be used for the repair program whereby individual tasks will be defined and a work package for each task, con-taining all pertinent information required to complete that task, will be completed, l

l 3.2.6.4 Post Shutdown Activities l

Following the shutdown of the unit, certain preparatory activities will be completed prior to the actual removal process. The following activi-ties are typical of those which will be performed; however, they are not necessarily listed in the order which they will be performed,

l. Establish appropriate valve lineups consistent with the require-ments of the repair efforts.

O 3-13A

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2. Place systems in the appropriate condition for long tem layup, l .

i.e. , approximately six (6) months. Loup procedures for primary and secondary systems during steam generator repair will be devel-oped. It is expected that wet layup using borated, demineralized water will be used for primary systems with hydrazine addition l and/or nitrogen blanketing, as appropriate, to prevent intrusion of

[ oxygen. Layup of secondary systems such as the feedwater system and condenser, is expected to be dry. These systems will be drained and dried with air in accordance with normal practice, i

3. Open equipment hatch and establish access control to work area.

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4. Remove reactor pressure vessel head and upper internals and store.

a i 5. Remove all fuel assemblies from the reactor vessel and store in

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i necessary. The existing polar crane will be used to make all major lifts.

17. Enlarge and/or reinforce equipment hatch area outside of the con-

l. tainment.

I 3.2.6.5 REMOVAL ACTIVITIES Having established appropriate access requirements, radiological controls, installation of temporary facilities to provide access to the work areas and for removal of debris and components, and the removal of insulation from the equipment, the actual removal process can commence.

A description of the basic removal process is given below; however, the l sequence of activities is not necessarily in order of implementation.

l f The description given below is applicable to one steam generator; however, the other steam generator will be removed in a similar manner.

The activities for both generator.s will be performed roughly simulta-neously; however, because of availability of the polar crane, the-commencement of the activities for each steam generator will be staggered, e.g., the removal process may not begin for steam generator number 2 until about a week after it begins for number 1.

Where cutting is required either thermal cutting techniques or mechanical techniques may be used.

1. Remove miscellaneous small piping, such as blowdown piping, and instruments and controls, such as level transmitters to facilitate removal of the steam generator.

2. Cut steam line piping at the steam nozzle on the upper shell and downstreams to allow a section of the piping to be removed so that the upper and lower shells can be lifted. The removed section will

, be marked for identification and stored for reuse. (See Figure 3-5) l l 3. Cut feedwater piping at its junction with the upper shell and lO l

i 3-15 L-. - - - - .

t upstream from the junction at'the concentric 16" x 18" 0.D. reducer to allow a section of the piping to be removed so that the upper and lower shell can be removed. (See Figure 3-6) l l

4. Cut and remove reactor coolant inlet and outlet piping. A section of the hot leg (inlet) piping (an elbow) will be removed by cutting the pipe at the steam generator nozzle and at an appropriate point

]

up stream of the nozzle on the hot leg piping. A larger section of cold leg (outlet)-piping, consisting of two elbows and two straight section, will be removed by cutting the pipe at the steam genera-

. tor nozzle and upstream of the reactor coolant pump. Figure 3-5

} schematically shows the portions of the reactor coolant piping l~ which will be removed.

5. Cut steam generator wrapper to facilitate lifting of the upper

assembly.

6. Cut steam generator shell at the transition cone to upper barrel girth weld leaving stock on the upper steam drum for final machining. The lower assembly will not be reused; however, the shell of the upper assembly will be used.

4 7. After removal of the upper assembly, it will be placed in conveni- -

I j ent location within the containment in the inverted position, i.e.,

steam nozzle down, where the moisture separation equipment, feed-l ring and other associated equipment will be removed and refurbished.

l

8. The steam generator lower assembly will be lifted from its

( supports by the polar crane. The polar crane will be attached to

( the lower shell by means of cables or straps attached to the two j lifting trunnions on either side of the steam generator. A l special lifting rig must be used to attach these cables to the crane hook. As shown in Figure 3-2, the steam generator will be l lifted straight up out of its supports, moved aside in the vertical position to a designated location on the operating floor,

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and lowered onto upending /downending skid in the vertical

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. - -- . _ . . = . _ = - _ - .

j 14. Reassemble the removable block shield wall for the steam generator p cubicles. -

1

15. Remove all temporary structures which were put inplace to facili-

tate the repair process.

16. Restore concrete structures which were chipped. ,

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3.2.6.7 POST INSTALLATION ACTIVITIES

, Following the completion of the major installation activities, it will i be necessary to restore the unit to a condition from which unit startup can connence, and to perform tests or inspections. The following are j typical activities which will be perfonned:

1 I

1. Hydrostatic tests and baseline inspections of piping or components j will be perfonned in accordance with Section XI of the ASE Code for piping or components affected by the steam generator repair.

These tydrostatic tests and inspections will assure the integrity I of the reactor coolant system. In addition to the hydrostatic tests following rep 0ir, the replacement steam generators will be hydrostatically tested on the primary side at the manufacturing l- facility. i l

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! Detailed procedures will be developed for the hydrostatic tests

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and inspections.

l- 2. Clean affected systems and work areas.

3. Install insulation.

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4. Remove scaffolding.

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5. Remove cavity cover.

I 6. Refuel the reactor.

3-19 l

1

7. Perform baseline inservice inspection as required on piping, equipment or components, including 100 percent eddy current 7-( inspections of steam generator tubing.

8. Remove all temporary structures and supports.

9. Install reactor internals, vessel head and other components.

3.2.6.7.1 Steam Generator Interior Inspection Cleaning criteria during fabrication and preparation for shipment of the steam generators are specified in WCAP-8370, Rev. 9A which includes criteria for both the primary and secondary sides of the steam genera-tors. The steam generators are sealed and protected against moisture during shipment and are maintained dry during installation. Thus, there is assurance that the steam generator surfaces will not be exposed to wet, oxygenated environments during installation and that primary and secondary surfaces are acceptably clean. Reopening the steam generator for an inspection following hydrostatic testing would expose surfaces to a wet, oxygenated environment which could result in fomation of iron oxides on carbon' steel steam generator surfaces in the secondary side.

As a consequence, the " metal clean" criterion implied by ANSI N45.2.1 -1973 paragraph 3.1.2(1) (as referenced in Regulatory Guide 1.37) probably could not be met on the secondary side of the steam generators following hydrostatic testing and cannot be met following a period of operation since the carbon steel surfaces would be covered with the normal film of magnetite and other iron oxides. Since primary surfaces

, are either Inconel or stainless steel and are verified clean following hydrotest at the manufacturing facility, there is no need for an inspec-tion following hydrotesting in the plant. Following the repair, final inspection and sean:h will be performed on steam generator secondary side. This search will be perfomed by inserting a fiberscope through L the steam generator handholes and conducting a 360 degree search of the annulus at the tubesheet. Any foreign objects which are judged to have the potential for steam generator tube damage will be removed.

O 3-19A

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'l j . tion of in containment work spaces are essential for minimizing exposure j while working in potentially high radiation areas. The computer aided design (CAD) computer system, modeling and scale drawings will be 'used

to confim access clearances for the movement of tools and equipment in j and out of containment. These techniques will minimize the potential  !

for unexpected delays in containment work and the associated radiation

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!- exposure. Work space envelopes will be studied to assure adequate space

! for the tooling designe'd to be used in the high radiation environment.

j 3.3.5.2 TEMPORARY SHIELDING Shielding will be used, as necessary, to reduce the dose rates from

other components such as the regenerative heat exchanger, RHR system j valves, and from temporary storage areas used for storage of contami-nated pieces of pipe, rags, and tools. Temporary shielding will be i~ used, as necessary, for the steam generator while it is being cut out of the reactor coolant loop and while the steam generator and reactor cool-ant pipe is moved out of the containment. The steam ganerator shell will also help shield the more contaminated parts of the stear, genera-  !'

i tors.  ;

The water level on the secondary side of the steam generator will also

be adjusted as required to provide shielding during cutting of the upper shell and feedwater and steam lines.

1 3.3.5.3 LOCAL DECONTAMINATION Local decontamination of the steam generators and reactor coolant piping O may be perfomed. Decontamination of the work areas will be perfomed periodically depending on the contamination levels. Paper and plastic ,

sheeting will be used to facilitate collection and cleanup of contamina-l tion.

No chemical decontamination of components is planned for the repair process. Thus, chemical decontamination fluids will not be introduced to systems or components affected by the repair. . Reactor coolant pipe O ends will be wiped with lint free rags dampened with demineralized water to reduce surface contamination. While details of weld preparation 3-26

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procedures have not bee.1 finalized, additional decontamination may be p perfonned using decontamination methods with abrasive material such as boric acid grit in a water slurry if determined to be required based l upon ALARA evaluation. Such decontamination would not introduce mate-rials deleterious to the reactor coolant system.

( 3.3.5.4 LOM BACKGROUND RADIATION WAITING AREAS Low background radiation waiting areas will be established where workers must wait between tasks. Special signs will be posted to designate v these areas. Signs will also be posted in high background radiation amas to warn personnel.

Health physics personnel will work with the job supervisors to assure that personnel r.ot required in the work area remain in the waiting ama.

3.3.5.5 RADIOLOGICAL PROTECTION PERSONNEL TRAINING As a minimum, personnel will be given radiological protection training as described in the Point Beach Nuclear Plant Health Physics Administra-tive Control Policies and Procedures Manual. This training consists of a radiation protection orientation given all personnel who work with radioactive materials prior to working unescorted in Radiation Con-trolled Areas and, as required, one of five additional health physics training courses given to auxiliary operator trainees, radiation control operator trainees, security guards, and plant supervisors. The orienta-tion program includes, but is not limited to instructions and demonstra-tions in Radiological Protection Program, Emergency Plan, fire alanns

(/ and response, and ALARA.

3.3.5.6 MOCK-UP TRAINING The extensive training of repair personnel on full-sized mock-ups has proven to be an effective method for minimizing personnel exposures during steam generator maintenance. Steam Generator Repair personnel will become thoroughly familiar with each tool that is designed and v tested in the mock-ups. As each tool reaches final design stages, field procedures will be written for qualifying the tool in the mock-up. Tool 3-27

designs will be modified and procedures will be updated for field imple-mentation based on lessons learned in mock-up testing.

In parallel to the field qualifications of the tooling, personnel will be trained and qualified to operate the equipment in field operations.

O v Technicians will be required to operate the equipment in simulated radiation conditions, dressed in the complete Anti-C clothing required for perfoming the repair in operating plant conditions. An estimate of time spent in the radiation environment will be used to detemine man-power mquirements and to establish administrative man-rem exposure limits for completing the repair. The training records of each techni-cian will be documented and forwarded to the field coordinator so that only qualified personnel are used in the steam generator replacement project. It has been the experience of Westinghouse over the years that

' ALARA issues involving radiation exposures to repair personnel are best addressed by using highly trained, experienced technicians to perform tasks in a high radiation environment.

3.3.5.7 REMOTE MONITORING SYSTEMS The utilization of remote monitoring systems will be used during the repair of steam generators. TV cameras will be used to monitor work both inside and outside of the steam generator cuoicle. The remote monitoring systems may be used for QA inspections as well as during

repair operations. The remote monitoring systems in combination with an audio communication system will be designed to minimize the time spent in the high radiation field.

3.3.6 MISCELLANEOUS WASTE DISPOSAL

+

3.3.6.1 CONCRETE DISPOSAL Approximately 15 cubic yards of concrete will be removed from the con-tainment internal walls and floors and will be disposed of. The majority of this concrete has an insignificant amount of transferable V contamination (transferable contamination is considered insignificant if it is less than 2200 dpm/100 cm2per 49 CFR 173.397) without surface 3-28

i decontamination. The concrete which is considered contaminated, (i) may be decontaminated prior to cutting by vacuuming and/or scrubbing with detergent and water to reduce the amount of transferable contamination to as low as is reasonably achievable or, (ii) appropriately packaged for shipment. Following removal from the containment, the concrete will be shipped as " low specific activity" (LSA) material to a licensed land O burial site.

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3.4 DISPOSITION OF STEAM GENERATOR LOWER ASSEMBLIES

( The lower assemblies to be removed from Point Beach Unit 1 represent the single largest source of solid radioactive waste to be disposed of dur-ing the repair effort. The disposal effort is independent of the repair and is evaluated on that basis.

J The primary side surfaces of the steam generators are covered by a tenacious film of deposited radioactive products containing primarily I cobalt isotopes. Based on actual Point Beach data provided in Section 5.2.2, it is estimated that at the time the lower assemblies are removed, each will contain approximately 300 curies of deposited gamma activity.

The removal procedures require that the steam generator lower assem-blies be sealed prior to movement out of the containment. Sealing will be accomplished by welding closure plates (covers) over the top of the lower assembly at the girth cut location and over the inlet and outlet reactor coolant nozzles and all other vessel penetrations. The sealing is accomplished to assure containment of the radioactivity within the lower assembly and to minimize the potential for radiation streaming from these penetrations. Thickness of the covers for the top of the lower assembly and nozzles will be such that structural integrity is assured and that radiation levels are reduced to acceptable levels.

Estimated thickness of the covers is approximately three (3) inches of steel or its shielding equivaient. The steam generator lower assembly primary and secondary sides will be drained. Essentially all the radio-activity is contained in the primary side of the steam generators which will be dry during the steam generator repair and storage. Thus, there is no need for further drying or gas blanketing of the steam generators.

3.4.1 OBJECTIVES OF HANDLING / DISPOSAL OPERATIONS O

v The objectives of handling / disposal operations are as follows:

A. To dispose of or store the lower assemblies safely and economically, and in accordance with applicable licensing requirements.

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B. To provide means to handle / dispose of the steam generator lower assemblies so that radiation exposures to personnel are as low as is reasonably achievable.

C. To minimize the potential for release of radioactivity to the envi-ronment so as to keep radiation exposure to the public as low as is reasonably achievable and within 10 CFR 20.

3.4.2 ONSITE STORAGE A temporary onsite storage building will be provided for the storage of the lower assemblies. It is expected that the lower assemblies will be stored in this buf1 ding until plant decommissioning. Prior to removal from the containment, the openings in the lower assemblies will be sealed by welding to prevent the release of radioactivity during l

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transfer and subsequent onsite storage. As discussed in Section 3.4.4, the only radiological consideration associated with storage is the direct radiation from the steam generators. Shielding will be provided to ensure acceptable radiation levels external to the storage facility.

Section 3.4.5 demonstrates that there are no safety concerns associated with onsite storage.

O Removed Reactor Coolant pipe and fittings will be capped with ends shielded, and stored with the Steam Generator lower assemblies in the temporary onsite storage building.

O Based on the above considerations, the required storage facility design criteria are:

A. Appropriate shielding for direct dose.

B. Provisions for periodic surveillance of the steam generator lower assemblies.

C. Provisions for preventing releases to the environment.

3.4.3 0FFSITE DISPOSAL Disposal of the steam generator lower assemblies at an offsite facility is not an available alternative at this time due to restrictions at existing disposal facilities. Therefore, detailed evalu&tions of alter-native offsite disposal methods have not been made specifically for the Point Beach lower assemblies. Estimates of occupational doses for various steam generator disposal alternatives have been made by Hoenes, et. al . (Reference 1). A summary of these estimates is provided in Table 3-2. With the exception of the immediate intact shipment alterna-tive, the long-term onsite storage alternatives provide the lowest esti-mated occupational exposures.

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i l 3.4.4 RADI0 ACTIVE RELEASES AND DOSE ASSESSENT ASSOCIATED WITH l

j ONSITE STORAGE i

4 As indicated in Section 3.4.2, prior to removal from the containment, ,

1 i the openings in the steam generator lower assemblies will be sealed to prevent the release of radioactivity during transfer and subsequent i

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onsite storage. Since the lower assemblies will be completely sealed, n there will be no airborne releases as a result of lower assembly onsite storage.

The only potentially radioactive liquid wastes associated with the m onsite storage of the lower assemblies are liquids collected in the temporary storage facility sump. If necessary, these can be processed through a radwaste evaporator and subsequently discharged, or solidi-fied, packaged, and shipped to a disposal site. Since the lower assem-blies will be sealed prior to transporting to the storage facility, it v is not expected that processing will be required.

As discussed in Section 3.4.5, the radioactivity within the steam gen-erators is immobile and the lower assemblies are stored in a closed facility. Thus, even if seal integrity were lost, releases to the envi-ronment are not likely. Nonetheless, a surveillance program will be implemented. Periodic area radiation surveys and monitoring will provide assurance that there are no releases of radioactivity to the environment.

G The only contribution, therefore, to the annual dose equivalent to any member of the public is from direct radiation emanating from the storage facility. The storage facility will be shielded, as required, in order to limit the dose rate at the outside limits of the storage facility to

<2.5 mr/hr. The resulting dose equivalent to an individual at the site boundary for a full year is estimated to be less than 0.01 mrem, which is insignificant. Furthemore, it is highly unlikely that an j p individual would be continuously exposed for a period of one year at the site boundary; therefore, the actual annual dose equivalent to any l'

individual at this location will be substantially lower than that given above.

.O V The ALARA approach to the steam generator replacement has been reviewed with regard to the estimated savings in personnel exposure which might be expected with decontamination of the steam generators prior to removal. The work sequence has been established such that the work in O close contact with the steam generators required to remove and transport them to storage locations has been minimized, thus minimizing the amount  ;

3-32

of personnel exposure. The highest source of radiation will result from the top of the steam generator tube bundle, and from the opening to the steam generator channel head. The removal procedures require welding shielded covers over the reactor coolant nozzles, a cover over the tran-sition cone, and on all sample lines, instrument penetrations, and the blowdown line. The intent of these covers is to use the shell as a shielded transport cask for removing the radioactive steam generator internals to the temporary storage area on site. This technique has been used in the past (Surry, Turkey Point) and is a proven method of removing the steam generators with minimal exposure'to repair personnel.

O Based on a review of state-of-the art decontamination processes, a chemical process would be required in order to effectively decontaminate the steam generator tubes. The oxide film on the internal steam genera-tor surfaces is tightly adherent and water washing or flushing processes are not expected to significantly reduce radiation levels in the steam generators. Radiation exposures would be received during the chemical decontamination, as well as in processing, packaging and handling large volumes of radioactive waste generated by tne processes. Occupational I exposure estimates for tasks carried out in preparation for steam gen-erator decontamination are listed in Table A.2 of NUREG/CR-1595, as follows:

Remove Manway Covers 20 man-rem Clean-up Manway Entries 80 man-rem l Remotely place inflatable plugs in coolant inlet and outlet 20 man-rem huJ Exposures associated with the chemical decontamination process are estimated to be an additional 24 man-rem in Table 8.1 of NUREG/CR-1595.

Thus, a total in the order of 140 to 150 man-rem per steam generator could be expected for chemical decontamination prior to removal.

The exposure savings which could be anticipated from chemical decon-tamination of the steam generators (including the steam generator tubing) is conservatively estimated to be 286 man-rem, assuming a dose reduction factor of 10 after decontamination. This was calculated by C

3-32A

considering the steps in Table 6-2 which would be affected by the decon-tamination process. The NUREG-CR-1595 data indicate that 140 to 150 O man-rem par steam generator or 280 to 300 man-rem total would be asso-ciated with the decontamination effort. Thus, no net exposure savings would be expected from chemical decontamination of the steam generators prior to removal. '

O The radiation dose to workers associated with transport and long-term storage is estimated to be only 10 man-rem per steam generator in NUREG-CR-1595. Assuming a decontamination factor of 10, this could be reduced to 1 man-rem per steam generator or a reduction of 18 man-rem total. This additional reduction in long term radiation exposure is insignificant when compared to the radiation exposure required to per- ,

form the decontamination.

3.4.5 ACCIDENT CONSIDERATIONS ASSOCIATED WITH ONSITE STORAGE The only potential accident consideration associated with steam gen-erator lower assembly storage is the release of radioactivity to the O

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3.5 PLANT SECURITY Specific plans for physical protection of Point Beach Nuclear Plant Units 1 and 2 during the steam generator repair will be addressed, as necessary, in a separate submittal to be withheld from public disclosure pursuant to 10 CFR Part 2, paragraph 2.790(d).

O 3.6 QUALITY ASSURANCE PROGRAM Wisconsin Electric Power Company (WE) has the overall responsibility for the Quality Assurance Program for installation of steam generators in accordance with Section 1.8 of the Final Safety Analysis Report. The regulatory guides applicable to the installation are those addressed in Section 1.8, Table 1.8-1 of the Final Safety Analysis Report, including additional Regulatory Guides 1.144 and 1.146.

The Quality Assurance Program used by Westinghouse Nuclear Services Integration Division (W NSID) for steam generator replacement activities consists of and includes the Westinghouse Nuclear Services Integration Division " Quality Assurance Program Plan", WCAP-9245, Rev. 6, June 1982, as supplemented with the "W NSID Steam Generator Replacement Services Quality Assurance Program Plan Supplement", Rev. O, November 1982.

WE will review and approve the Westinghouse Quality Assurance Program in accordance with Section 1.8 of the Final Safety Analysis Report and as required by 10CFR50, Appendix B. WE QA personnel will audit and provide surveillance tu the extent necessary to assure that all activities are conducted in accordance with the applicable codes, standards and regula-tions and in accordance with the WE QA Program.

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4.0 RETURN-TO-SERVICE TESTING 1;

The repair of steam generators in Unit 1 will have minimal impact on I existing equipment and structures except for the steam generators and associated piping and instrumentation in containment. Therefore pre-operational testing requirements will be similar to those for a normal

,' refueling outage.

Upon completion of the steam generator repair, piping systems will be hydrostatically tested in accordance with applicable codes. Instrumen-tation and electrical equipment which was removed or relocated during the steam generator repair will be reinstalled and tested using normal maintenance procedures and Technical Specification requirements, as necessary. Other systems and components will be tested in accordance with Technical Specification requirements for testing prior to return to I'

power from a nomal refueling.

Testing and inspection of the repaired steam generators and other affected systems or components will include the following:

1. Steam generator thermal .perfomance tests to verify the themal performance parameters specified for the repaired steam generators.

3

2. Steam generator moisture carryover tests to verify that moisture i carryover in the steam is within design values.

j 3. Calorimetric tests to verify adequate reactor coolant flow in accor-dance with Technical Specifications.

, s.

O 4. Inspection of equipment supports affected by repair activities in both hot and cold conditions.

1

5. Eddy current examination of steam generator tubes.

These tests and inspections are not significantly different from tests and inspections performed following nomal refueling or maintenance and provide adequate assurance that operation of the repaired steam genera-l tors will not affect the health and safety of the public.

! 4-1

, . - - , - . - . . - - - - - - . ~ . . - - . - - - . . - . . . . , , , _ _

d 5.2.2.5 Comparison with Observed Radioactive Liquid Releases During Nomal Operation Estimated radioactive liquid releases during the repair effort are compared with the observed liquid waste releases during the year 1981 in Tab le 5.2-11. The estimated total radioactive liquid release per unit (excluding tritium and dissolved gases) during the repair effort is seen to be about 42 percent of the observed total liquid waste release per unit (excluding tritie, and dissolved gases) during 1981. The estimated tritium release per unit during the replacement effort is about 38 O

V percent of the observed tritium release per unit during 1981.

5.2.3 Spent Fuel Pool Cooling Considerations As described in Section 9.3.1 of the FSAR, the design basis for the spent fuel pool cooling system assumes a full core unload for inservice inspection or_ maintenance of reactor coolant system components with spent fuel assenblies stored in all other spent fuel rack storage loca-tions. Thus, the spent fuel pool cooling system is adequate to accom-modate the core unload required for repair of the Unit 1 steam generators.

The core unload required for steam generator repair would not be expected to result in radioactivity concentrations significantly dif-ferent from those resulting from storage of spent fuel, refueling opera-i tions or unloading of the core for routine inservice inspection. Expe-

! rience during past core unloads and normal refuelings at Point Beach has

> indicated that the spent fuel pool purification system is adequate to maintain spent fuel pool radioactivity concentrations at acceptably low levels. Thus, the spent fuel pool purification system is adequate for the core unload required for steam generator repair.

5.2.4 References for Section 5.2

1. Hoenes, G.R., M. A. Mueller, W. D. McConnack,1980, Radiological Assessment of Steam Generator Removal and Replacement: Update and Revision, NUREG-CR-1595 (PNL-3454), U.S. NRC, Washington, D.C.

1 I

5-29

2. Virginia Electric and Power Company,1979, Steam Generator Repair Program for the Surry Power Station Unit No. 2 - Final Report (Progress Report - No. 6) for the Period February 3,1979 through December 31, 1979, NRC Docket Nunters 50-280 and 50-281, Washington, D.C.

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performing the work, e.g. temporary shielding. Special emphasis has been placed on the engineering and planning prior to issuing " work pack-ages" to assure that ALARA objectives are incorporated.

1 The steam generator project organization reflects a commitment to ALARA objectives. A full time engineer knowledgeable with health physics I practices is assigned to the headquarters organization to assist in the i

planning phase. During the engineering and planning phase, he is assisting in establishing basic criteria for implementing ALARA objec-tives, as well as working with the engineers, planners and consultants in establishing specific requirements. During the actual work a full 4 time Health Physics Director, knowledgeable with health physics prac-tice, will implement all health physics activities, through the Health

. Physics Manager, Health Physics Shift Coordinators, and Health Physics

Technicians as described in Figure 6-1. The Health Physics Director

! will have line responsibility through the Site Manager and the Program Engineering Manager under Headquarters Engineering to the WE Special ,

Projects Administrator. Interface with WE health physics personnel is

, as described in Figure 6-2.

iO

Health physics procedures written for the steam generator project will j be utilized. These procedures provide instructions for HP related items and implement applicable Point Beach Health Physics Procedures and Policies. The Health Physics Director will be responsible for assuring i that an effective measurement system is established, that results are

, reviewed with the WE Health Physicist, and that corrective actions are taken when attainment of the specific objectives appear to be com-promised. The appropriate resources needed to achieve ALARA goals and objectives will be provided. The Health Physics Director and Health

, Physics Shift Coordinators will be able to call upon the headquarters l

organization as well as consultants for support.

l

! The basic responsibilities of the Health Physics Director for the

! replacement activities includes:

l- A. Participating in design reviews for facilities and equipment that can affect potential radiation exposures; 4

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steam g2nerator project will receive instructions and training in expo-l[

!- sure control and emergency procedures. All personnel involved in steam

• generator repair activities whose duties require (1) working with radio-active materials, (2) entering radiation areas, or (3) directing the

j activities of others who work with radioactive materials or enter radia-tion amas will mceive training. The training program includes suf-li.

i

ficient instruction in the biological effects of exposure to radiation j

3 to pemit the individuals receiving the instruction to understand and i evaluate the significance of radiation ' doses in tems of the potential

risks. The training program also includes instruction on radiation protection rules for the plant and the applicable- federal regulations.

i' The scope of this training program is responsive to the recommendations

! in Regulatory Guides 8.13, 8.27, and 8.29.

1 Additional radiation protection aspects which are responsive to Regula-tory Guide 8.27 recommendations for use of mock-up and work tasks train-ing are outlined in Section 3.3.5.6. This training will be performed by j, qualified personnel experienced in the tasks to be perfomed to assure that trainees are able to perfom the work in a satisfactory and safe l'

manner. Qualification sign-off sheets and records of training attendance will be maintained.

1i 4

!' It is planned to utilize experienced and trained personnel to implement l I

, the replacement program. The use of highly skilled craf t labor should

!' permit tasks to be perfomed reliably and more effeciently. Specific training sessions will be held for tasks unique to the replacement

!l

)' activi ties.

LO 6.4 ENGINEERIhG AND DESIGN REVIEWS I! The overall steam generator replacement program will be implemented by I

the use of " work packages" that are amendle to efficient and timely i

review. These packages contain all the infomation required to imple-fj ment a specific task associated with the project, e.g. cutting the steam i'

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generator. Each of these packages is subject to intensive review, including operation, maintenance, construction, quality assurance, health physics, and engineering personnel. The coordination in the

various groups is the responsibility of the cognizant engineer. This e

coordinated effort by these individuals ensures that the objectives of the ALARA Program are achieved.

To the extent possible, the repair activities reflect considerations of personnel required to perfor1n maintenance and inservice inspection oper-ations that could lead to substantial personnel exposures. Speci-

,V fications for repair equipment reflect the objectives of ALARA as shown by the following examples:

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TABLE 6-1 (Continued) 5 of 42 4

R.G. 8.8 C0t#ENT STEAM GENERATOR REPLACEENT PROGRAM PROVISIONS

, rinths for access. However, such labyrinths The steam generator primary side manway cover j or other design features of the cubicle will remain in place as long as practical >

j should permit the components to be removea during the repair process to eliminate readily from the cubicle for repair or re- streaming. When the manway covers are removed

placement where such is work expected or antici- to install venti 11ation or reactor coolant

! pated. Single-scatter labyrinths may be pipe dams, administrative controls will be i inadequate if the cubicle contains a sub- implemented to minimize personnel exposure

! stantial radiation source. due to channel head streaming.

(5) Streaming of radiation into accessible areas 5. Streaming of radiation will be minimized through penetrations for pipes, ducts, and by installing shielding, such as plugs in other shield discontinuities can be reduced open ended pipe lines following cutting.

(a) by means of layouts that prevent sub-stantial radiation sources within the shield from being aligned with the penetra-tions or (b) by using " shadow" shields such as shields of limited size that attenuate

the direct radiatior, component. Streaming also can occur through roofs or floors un-less adequate shielding encloses the source from all directions.

(6) The exposure of station personnel to radiation 6. Not specifically applicable to repair from pipes carrying radioactive material can program.

be reduced by means of shielded chases.

(7) Design features that permit the rapid removal 7. The insulation presently installed on the

and reassembly of shelding, insulation, and steam generator and certain portions of other material from equipment that must be the piping connected thereto will not be 4 inspected or serviced periodically can reduce be reused. New reflective type insula-the exposure of station personnel perfonning tion will be designed to provide qttick

, these activities. and easy access to areas subject to in-l service inspection.

l o o o o o a: a TABLE 6-1 (Continued) 28 of 42 R.G. 8.8 C0 K NT STEAM GEERATOR REPLACEMENT PROGRAM PROVISIONS .

l (1) A staff member who is a specialist in 1. A Health Physics Manager will be part of the 4

radiation protection can be assigned the project staff. He will be a taff member responsibility for contributing to and qualified and experienced in diation pro-coordinating ALARA efforts in support of tection and.will work full ti to assist

uperations that could result in substantial in preparing the" work packages . During individual and collective dose levels. the actual replacement activities, he will be responsible for all radiation protection activities, including dose control and moni-toring, issuance of radiation work permits, etc. The specific responsibilities are identified in the position description for the Health Physics Manager.

(2) To provide the bases for planning the acti- As part of the preplanning activities vity, surveys can be performed to ascertain historic survey data and experience has

infomation with respect to radiation, con- been reviewed. Supplemental surveys will tamination, airborne radioactive material, have also been perfomed. At the beginning and mechanical difficulties that might be of the outage it is planned to make compre-encountered while performing services. hensive surveys prior to the commencement

. of work and to conduct them periodically I thereafter. Previous operating experience l has been reviewed regarding mechanical difficulties and this information is j being incorporated into the work pack-ages. Since the original installation of i the steam generators is very similar to that planned for replacement, this knowledge has been utilized, such as use of photographs, review of records, etc.

}

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_ _ _- _~

i O O O O O O O l TABLE 6-1 (Continued)- 32 of 42 i

R.G. 8.8 COMENT STEAM GEERATOR REPLACEENT PROGRAM PROVISIONS l

i i can reduce the potential for such occurrences

, end enhance the capability for coping with the situations expeditiously if they occur.

! (10) Portable or temporary shielding can reduce 10. Temporary shielding, e.g. lead blankets l dose rate levels near " hot spots" and in will be used to minimize dose rate levels.

the general area where the work is to be For instance, in the cutting and reweld-j performed. ing of the reactor coolant piping, shield-ing will be used to isolate those portions of the system which are contaminated.

j (11) Portable or temporary ventilation systems 11. Temporary ventilation systems will be j or contamination enclosures and expendable used for certain work tasks, e.g. cutting

! floor coverings can control the spread of of reactor coolant piping. Where appro-i contamination and limit the intake by priate, coverings will be used to minimize l workers through inhalation. spread of contamination.

} (12) " Dry runs" on mock-up equipment can be 12. Dry runs and mock-up equipment will be used for training personnel and testing useful for training personnel, identifying problems that can be encountered in the' equipment. For example, a mock-up of j actual task situation, and selecting and the channel head and the transition cone ,

qualifying special tools and procedures areas will be used to simulate work in the i to reduce potential exposures of station high exposure areas associated with the I personnel, welding of the generator. The mock-ups will

} simulate the space constraints associated 1

with the planned work. The actual l

equipment to be used at the site will be

used for training. Proven tools and equipment will be used extensively.

1 i l

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O O O O O O O TABLE 6-1 (Continued) 35 of 42 R.G. 8.8 CO E NT STEAM GEERATOR REPLACEMENT PROGRAM PROVISIONS I b. Operations During operations in radiation areas, adequate supervision and radiation protection surveil- '

lance should be provided to ensure that the appropriate procedures are followed, that

planned precautions are observed, and that

~

, all potential radiation hazards that might develop or that might be recognized during the operation are addressed in a timely and appropriate manner.

(1) Assigning a health physics (i.e., radiation 1. There will be an adequate number of health safety or radiation protection) technician physics personnel assigned to each shift.

the responsibility for providing radiation The duties of these individuals will be protection surveillance for each shift related to the steam generator replacement operating crew can help ensure adequate activities. Other members of the health radiation protection surveillance. physics staff who are assigned to the operating unit would be available in

, unusual situations.

(2) Personnel monitoring equipment such as 2. Direct reading dosimeters and TLD's will direct-reading dosimeters, alarming dosi- be used to determine doses to individuals meters, and personnel dose rate meters can be used to provide early evaluation of doses to individuals and the assignment i of those doses to specific operations.

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7.0 ENVIROMENTAL ASPECTS OF THE REPAIR 7.1 GENERAL This section evaluates environmental effects relevant to the steam gen-i erator repair effort and demonstrates that no significant environmental l

effects are associated with the repair activities. Any minor environ-mental impacts are expected to be temporary and controllable by the use

of standard construction practices. The site preparation, construction, and repair activities will be carried out in conformance with local, state and federal regulations.

7.2 RESOURCES C0p#1ITTED 7.2.1 Non-Recyclable Building Materials Housekeeping operations for all construction areas will be performed throughout the constuction period. Construction wastes will be separated into salvageable and non-salvageable materials. Salvageable materials such as lumber and scrap metal will be sold to salvage con-tractors. Non-salvageable materials will be disposed 'of by a licensed contractor.

l All required fuels, oils, and chemicals will be handled, stored and disposed of in accordance with applicable Wisconsin Administrative Code regulations. Any spills will be cleaned up quickly and any contaminated materials properly disposed.

7.2.2 Land Resources The repair effort will have minimal impact on existing site layout and plant facilities. Three new facilities, a 9,360-square-foot operations building and a 9,000-square-foot temporary steam generator storage

+

building, and a 5,400 square-foot shop building will be constructed in the previously modified area adjacent to the plant to the north. If O

i 7-1

. . _ _ _ - _ _ . - . . =- - . .. . .

required, a 's:cond 6,480 square-foot operations building may also be ,

j constructed. A 10,800-square-foot access structure will be constructed Y adjacent to the Unit 1 containment facade. Only about 3.7 acres of land V will be required for construction activities. Two parking lots will require about three acres. The temporary steam generator storage building, operations buildings and shop building will require about 0.7

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acres. Figure 7-1 shows the proposed plot plan for the repair activity l

facilities and their location with respect to existing plant buildings

and facilities.

No historical, cultural, archeological sites, or natural landmarks or access thereto will be affected by the proposed construction activities.

l

! Erosion and runoff control measures include: 1) limiting site grading ,

' and surface disturbance to the minimum area practicable, and 2) covering construction roads and parking areas with gravel. In addition, an approximately 210-foot-wide vegetated area will serve to filter sedi-ments from any parking area runoff before reaching Lake Michigan. Fol-

) lowing completion of the construction and repair activities, remaining disturbed areas around the buildings will be seeded to return to grass

' Cover. ,

The replacement steam generator lower assemblies will be shipped by rail to Kewaunee, Wisconsin and will be delivered to the plant site by over-1 land transporter. Thus, reconstruction of a barge slip at the site is

} not necessary.

i The construction area to the north of the plant is utilized minimally by the species of fauna known to inhabit the plant site. It is anticipated f' that whatever species nomally use the habitat in the construction area l

!, will move to and use other adjacent areas of the plant site once con-

'. struction activities begin, and other existing populations of wildlife on the plant site will avoid the area until construction activities are

completed. Approximately 3.7 acres will be lost.

!O 1

7-2

l, No displactment of wildlife from other areas of the plant site (i.e.,

wood lots, small ponds and stream course areas) due to increased levels

, of human activity and noise associated with the construction activities is expected to occur.

d No-impact on rare and endangered species is expected, f- ,

i 7.2.3 Water. Resourtes

, Potable water for the operations buildings and shop building will be supplied from an existing well located adjacent to the existing con-

, O struction building. An estimated 5,000 to 10,000 gallons of water per i day will be used during construction. No impact on the existing ground-

. water aquifier is anticipated.  !

i No groundwater impacts due to construction activities are anti.cipated.

i No dewatering of the site is required. Holding tanks will be used for

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handling sanitary facilities' wastewater and no grVundwater discharges j will occur.

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- 7. 3 WASTE WATER 1

i 7.3.1 Sanitary Facilities l

. Sanitary wastes will not be discharged on site. During site preparation j and early stages of construction, portable sanitary facilities will be l utilized. A holding tank will be used to collect sanitary waste from f1 the existing construction building and the operations building.

l Sanitary wastes from the containment access building will be processed l$s in the existing onsite sewerage treatment plant. Wastes will be removed I

LO LO 7-3

. . _ . _ _ _ , . _ ~ . . . . - _ . . - _ . - _ . _ _ . _ _ _ _ _ . _ - _ _ _ _ - _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ .

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by a lic:nsed contractor. The holding tank for the existing construc-I tion building' and operations building will remain after construction is lf completed. - The existing plant sewage treatment system has adequate capacity to process the additional wasteloads without modification.

41 -

}i ~ 7.3.2 Laundering Operations F' O Laundry waste water generated during the repair activities will origi-

); nate from facilities adapted for the proposed effort. If required, j; laundry waste water will be directed to the liquid radwaste system for

! processing (see Subsection 3.3.6.3). Additional infonnation on the expected quality and quantity of laundry waste water from T.he steam generator repair program is provided in Section 5.2.2.4.

7.4 CONSTRUCTION 4e The construction activities associated with the steam generator repair I .are not unique. The use of standard construction practices will result i i in effective control of the anticipated impacts.

UO

. -7.4.1 Noise -

l-i' l Noise levels in the construction area will be typical of- those asso-l ciated with the operation of site clearing equipment such as tractors, l bulldozers, front-end loaders, scrapers, trucks, and other construction equipment such as cranes, air compressors, and metal-cutting devices.

Typical sound pressure levels will range from 75 dBA to 110 dBA in the

!. vicinity of the equipment. No use of explosives will be required.

Standard noise control procedures, including the use of muffled equip-f ment, will be used to reduce noise levels. Occupational Safety and l

i Health Administration Standards (OSHA) will be followed to protect l; personnel located on site. Noise inputs are expected to be confined to the construction area.

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7.4.2 Dust-O oust wiii be created durine site eradine and b, mevement e< ve84cles on the unpaved construction areas. The primary means.of control will be _

e periodic sprinkling of the unpaved areas using water sprinkler trucks if the need is indicated by visual inspection. .The parking lot areas will'

~

-be covered with stone or gravel. Only about 3.7 acres will require grading. Following covering of the parking lots only 0.7 acres of land for the new buildings will remain in a disturbed state. The building sites will be open for a short period of time until the foundations are 3

poured. Dust is not expected to be a problem and any minor. impacts will  ;

be confined to the isenediate areas near where the site surface is dis- ,

i turbed.- t 7.4.3 Open Burning There will be no open burning.

7.5 RADIOLOGICAL ASPECTS The estimated releases of radioactive airborne and liquid effluents during the repair. effort are found to be much smaller than observed effluent releases for the operating plant during 1981. The comparison l is shown in Section 5.2.2. The radioactive effluent release points -

-during steam generator repair activities will be the same as during normal plant operations. ,

L Since releases of radioactive effluents during the repair program will be a small fraction of normal operating plant releases and their poten-tial exposure pathways will be the same as for the existing plant, the radiological impact of these releases is insignificant. These releases- '

will be monitored in accordance with the existing Point Beach environ-mental monitoring program.

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. 7.6 RETURN TO OPERATION

. 7.6.1 Water Use An existing well located in the area north of the plant provides a potable water source for the construction, operations and shop buildings. Maximum daily water use from the well is estimated to be about 5,000 to 10,000 gallons. No impact on the existing groundwater aquifier is anticipated.

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231 w. s a m aan,P.o. set sees. meLiaWie, el asset November 22, 1982 Mr. B. R. Denton, Director Office of Nuclear Reactor Regulation

, U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 .

Attention: Mr. R. A. Clark, Chief Operating Reactor Branch 3 Gentlement DOCKET NO. 50-266

, . RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION STEAM GENERATOR REPAIR POINT BEACH NUCLEAR PLANT, UNIT 1 I

Attached is our response to the request for additional informatiion, as requested by Mr. T. Colburn of your staff, regarding the Point Beach Nuclear Plant, Unit 1, steam generator repair.

Should you have further questions, please contact us.

Very truly yours, e

WL c .., ,

Assistant Vice' President C. W. Pay Attactunent copy to ASLB service List NRC Resident Inspector Blind copies to Messrs. R. W. Britt, Sol Burstein g . H. Gorske/

A. W. Pinke, D. K. Porter, J. J. Each, D. B. Tate, Gerald Charnoff, J. Woeber (W), B. Spezialetti (W)

ATTACHMENT 1 LICENSEE'S RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO

'O POINT BEACH NUCLEAR PLANT UNIT 1 STEAM CEATRATOR REPAIR QUESTION 1 O With regard to heavy load handling dur bg the proposed steam generator replacement, we will need the following commitment in accordance with ANSI 830.2.0, " Overhead and Gantry Cranes", as referenced in NUREG-0612. " Control of Heavy Loads at ~

Nuclear Power Plants".

i

a. Maintain complete records on the polar crane of all heavy loads handling operations during the repair phase.

b. Perform the indicated inspections and load tests on the polar crane after completion of the repair phase and prior to returning the crane to its nor-mal service.

RESPONSE: ,

s. Special heavy lifts during steam generator repair, as defined by ANSI 830.2.0-1976, will meet the requirements of Section 2-3.2.1 which details the require-ments for the handling of special heavy lifts. Additionally, normal indus-trial and construction standards for the handling of high value equipment will be observed. As stated in Section 3.2.6.2 of the Repair Report, the handling of heavy loads in the containment has no safety impact on plant operation, since all fuel will be removed from the reactor during the steam generator repair. Therefore, it is considered that the detailed requirements and procedures which have been implemented in response to NUREG-0612 are not applicable to the steam generator repair activities.

Records of special heavy lifts will be maintained in project records as part of the Control Work Packages discussed in Section 3.2.6.3 and Section 6, paragraph 6.4, of the Repair Report.

O l . - - - ..-.. _

b. The inspections indicated in ANSI B30.2.0-1976, Chapter 2-2, will be per-formed after completion of the steam generator repair and prior to return-ing the crane to its normal service.

l The existing overhead trolley will not be used for special heavy lifts dur-i ing the steam generator repair and will not be modified in any manner. Thus,

- a load test for the overhead trolley will not be required. The design stres-

~

ses for the polar crane girders will not be exceeded utilizing a center post.

Presently, no structural modifications to the bridge are anticipated. There-fore, a load test for the bridge will not be required. In the event that bridge modifications are required, the lifts performed during the repair project will constitute an adequate load test of the modifed bridge in lieu of the load test required by ANSI B30.2.0-1976, paragraph 2-2.2.2.

f QUESTION 2 ,

Verify that the steam generator repair or replacement will not introduce a feedring 4

design which will create a potential for steam generator waterhammer or commit to performance of waterhammer tests as a part of the return-to-service testing program l by automatic initiation of auxiliary feedwater in accordance with BTP-ASB-10-2.

i

RESPONSE

Steam generator waterhammer is believed to be caused by the rapid condensation of steam within the feedwater line. The presence of steam in the feedwater line can occur during hot standby operation when the steam generator water level drops

~

below the feedring elevation, allowing the feedring to drain and steam to enter the piping. The subsequent addition of cold auxiliary feedwater can cause the '

steam to condense rapidly, resulting in waterhammer.

The Model 44F replacement steam generators utilize a feedring designed to lieft drainage during such operations by the use of top discharge J-nozzles and a e .

welded feedwater nozzle themal liner. This configuration has been demonstrated by field experience to effectively limit the potential for waterhammer; examples include steam generators backfit with J-nozzles, operating experience with the Model 51F steam generators at Surry 1 and 2 and operating experience with the Model 44F steam generators at Turkey Point 3 and 4.

In addition, the feedwater piping configuration at Point Beach Nuclear Plant

-^ includes feedwater check valves located close to the steam generator feedwater nozzles. This configuration minimizes the potential for significant quantities of steam in the feedwater piping and thus minimizes the potential for waterhammer to occur.

Based upon'the design of the feedring for the replacement steam generators, field experience with steam generators ' utilizing this design, and experience with the piping configuration at Point Beach, the performance of waterhammer tests fol-lowing repair of the steam generators is not necessary.

QUESTION 3 Loose parts and foreign objects left inside steam generators have been identified as the cause of at least two steam generator tube rupture events. Recent inspec-l tions have found a variety of foreign objects in the secondary side of steam gen-erators. Please describe the precautions and prograas that will be implemented to preclude the incorporation of foreign objects into the steam generators during the current repair procedures as well as future surveillance methods to be imple-O mented.

RESPONSE

Procedures will be implemented to preclude the introduction of foreign objects C into the steam generators during the repair. These procedures include a combina-tion of physical barriers and administrative controls. Physical barriers will be specified as part of the Control Work Packages, consistent with the work to be O

ll performed. Physical barriers such as herculite, metal plates, and decking will be used in work aireas, as appropriate.

Ahinistrative procedures will include procedures for personnel access control, tool control and log-in and log-out procedures. The adainistrative controls will include such requirements as lacyards on small equipment and personal items, cod design features such as lock wires on equipment to prevent loss of material in the steam generators. Following the repair, final inspection and search will 4

be performed on steam generator secondary side. This search will be performed

< by inserting c fiberscope through the steam generator handholes and conducting a 360 degree search of the annulus at the tubesheet. Any foreign objects which cre judged to have the potential for steam generator tube damage, and which are cccessable, will be removed.

V Surveillance of the steam generators during subsequent operation could include periodic inspections of the tube bundles using fibre optic or television tech-i niques during refueling shutdowns, continuous monitoring via loose parts acnitoring systems, or a combination of these. It is our understanding that recommendations for loose parts surveillance are currently being developed by the NRC staff.and will be made available for comment prior to implementation. Loose parts surveillance programs during subsequent op2 ration will be developed following further definition of these recommendations. In any event, loose parts surveil-

lance programs during operation, if required, would be implemented whether or not the steam generators had been replaced.

QUESTION 4

- The Licensee should provide information demonstrating that decontamination of the steam generatcrs would result in no significant dose reduction to plant O, personnel, as stated in Table 6-1, page 14 of 42, item 4.

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EESPONSE:

The ALARA approach to the steam generator. replacement at Point Beach Unit I has L

been reviewed with regard to the estimated savings in personnel exposure which might be expected with decontamination of the steam generators prior to removal. ,

The work sequence has been established such that v.he work in close contact with 1 the steam generators required to remove and transport them to st.orage locations has been minimized, thus minimizing the amount of personnel exposure. The highest source of radiation will result from the tcp of the steam generator tube bundle, and from the opening to the steam generator channel head. The removal

, procedures require welding shielded covers over the reactor coolant nozzles, a cover ov,er the U-bend transition cone, and on all sample lines, instrument t

, penetrat' ions, and the blowdown line. The 1.ntent cf these cones is to use the shell as' a shielded transport cask for removing the radioactive steam generator internal's to the tempora:y storage area on site. This technique has been used in the past (Surry, Turkey Point) and is a proven method of removing the steam generators with minimal exposure to repair personnel.

Based on a review of state-of-the art contamination processes, a chemical process would be required in order to effectively decontaminate the steam generator tubes. The oxide film on the internal steam generator surfaces is tightly adherent and water washing or flushing processes are not expected to signifi-cantly reduce radiation levels in the staan. generators. Radiation exposures would be received during the etemical decontamination, as well as in processing packaging and handling large volumes of radioactive waste generated by the processes. Occupational exposure estimates for tasks carried out in preparation for steam generator decontamination are listed in Table A.2 of NUREG/CR-1595, as follows:

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i Remove Manway Covers 20 man-ren Clean-up Manway Entries 80 man-rom i

Remotely place inflatable plugs in coolant inlet and outlet 20 Man-rom Exposures associated with the chemical decontamination process are estimated to i

be an additional 24 man-res in Table 8.1 of NUREG/CR-1595. Thus, a total in the order of 140 to 150 man-res per steam generator could be expected for chemical decontamination prior to removal.

~

j S

The exposure savings which cculd be anticipated from chemical decontami nati on of 1

j the steam generators (including the steam generator tubing) is conservatively 4

estimated to be 286 man-rem, assuming a dose reduction factor of 10 after decon-tamination This was calculated by considering the steps in Table 6-2 of The Point Beach Unit 1 Steam Generator Repair Report which would be affected by the

decontamination process. The NUREG-CR-1595 data indicate that 140 to 150 man-rem per steani generator or 280 to 300 man-res total would be associated with the decontamination effort. Thus, no net expo m e shs hgs would be expected from

! chemical decontamination of the steam generators prior to removal. 4 The radiation dose to workers associated with transport and long-tem storage is estimated to be only 10 man-res per steam generator in NUREG-CR-1595. Assuming a decontamination factor of 10, this could be reduced'to 1 man-res per steam

• O generator or a reduction of 18 man-rem total. This additional reduction in l

l long tem radiation exposure is insignificant when compared to the radiation r

> exposure required to perform the decontamination.

i

_. _ _-_ _ _ ._ _ _ _ -____ _ _______.- _ - _ . _ . _ _ _ _ , , _ . _ , . . . _ . _ , . . - =

o decontamination factor of 10, this could be reduced to 1 man-res per steam generator or a reduction of 18 san-rem total. This additional reduction in long tem radiation exposure is insignificant when compared to the radiation

, exposure required to perform the decontamination.

QUESTION 5 The Licensee should provide a commitment to train personnel in accordance with Regulatory Guides 8.13, 8.27, and 8.29 or submit acceptable alternatives.

RESPONSE

All contractor personnel requiring controlled side entry into the Point Beach facility during steam generator repair will receive radiological protection training prior to working unescorted in Rt.diation Controlled Areas as outlined in Section 3.3.5.4 of the Repair Report. The orientation program is conducted by qualified training staff using written training saterials, oral presentation, video tape presentation, question-answer session, and a written examination to verify adequate comprehension of the information presented. The scope of this orientation program is responsive to the recommendations in Regulatory Guides 8.13, 8.27, and 8.29.

Additional radiation protection aspects which are responsive to Regulatory Guide 8.27 recommendations for use of sock-up and work tasks training are out-lined in Sections 3.3.5.5 and 6.3 of the Repair Report. This training will be performed by qualified personnel experienced in the tasks to be performed to assure that trainees are able to perform the work in a satisfactory and safe manner. Qualification' sign-off sheets and records of training attendance will be maintained.

O The requirements for training will be included in the Project Procedures Manual.

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OUESTION 6 .

Section ?.1.4 should include a commitment to the following Regulatory Guides O- or the Licensee should justify their exclusion: Regulatory Guides 1.39, 1.58,-1.88, 1.94, 1.44, and 1.46.

• j RESPONSE:

These commitments have been.made previously. Licensee's October 27, 1982 letter included commitments to the guidance of Regulatory Guides 1.58, 1.88,

}' 1.144, and 1.146 as applicable to the design and fabrication of the replacement steam generators and Regulatory Guides 1.39, 1.58, 1.88, 1.94, 1.144, and 1.146 as applicable to the installation of the steam generators with the clarifications noted in Attachment I to that letter. .

QUESTION 7 ,

As principal contractor of WE for the steam generator repair, Westinghouse l, should have its QA program descriptions (s) submitted to and found acceptable by the NRC. WCAP-8370, which describes the QA program for the Westinghouse Nuclear Technology and Nuclear Components Divisions, has been so processed.-

1 It should be referenced in Section 3.6.4 of the WE report as it is already l referenced in Section 3.6.3. WCAP-9245, which describes the QA program for the Westinghouse Nuclear Service Division, should be submitted for NRC Staff review.

! RESPONSE:

Licensee's October 27, 1982 letter indicated a commitment to reference WCAP-8370 t

in Section 3.6.4 of the Repair Report. WCAP-9245 (Revision 6) was also submitted f or information as Attachment 2 to that letter. A supplement to WCAP-9245 (Revi-O sion 6) is being prepared which will specifically address the Point Beach steam generator repair activities. This supplement will be submitted for NRC Staff information when it becomes available. .

O GUESTION 8 Will there be any change in the ass-rt of demineralizer waste discharged l

to Lah Michigan as a result of replscing the steam generators?

! l lP

j

RESPONSE

i O Domineralized water from the makeup water treatment system is supplied to various

( systems at Point Beach Nuclear Plant and is supplied to the secondary systems to i replace water removed via steam generator blowdown. It is expected that steam generator blowdown flow rates following repair of the steam generators will be a

similar to those for the existing steam generators. Thus, it is not expected that'significant changes will occur in the total quantities of neutralized regenerant waste released to take Michigan from the makeup water system.

f GUESTION 9

-Are any changes to the WPDES permit anticipated as a result of this action?

I RESPONSE: .

The repair of steam generators will not significantly affect waste releases to Lake Michigan, and it is not expected that changes to the WPDES permit will be necessary.

j QUESTION 10 -

Will the parking lot and layout areas be on previously distrubed areas?

l

RESPONSE

l Parking and laydown areas described in Section 7 of the Repair Report are located in areas which were previously disturbed for parking, laydown and con-crete batch plant operations during construction of Point Beach Nuclear Plant.

As stated in Section 7.2.2 of the Repair Report, these areas are utilized min-inally by species of fauna known to inhabit the site.

4 QUESTION 11 cD ' * " " ' '" ' "' ' '"'

"If so, please provide a copy.

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RESPONSE

It is presently planned to ship the re' placement steam generatore by rail to Kewaunee, Wisconsin and by road transport from Kewaunee to the site. Thus, reconstruction of the barge slip at the site is not necessary and a Corps of Engineers permit is not required.

1 f

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WISCONSIN Electnc macower O zal w. sicmCAN, P.O. BOK 2046, MILWAUREE, WI 53201 December 22, 1982 Mr. H. R. Denton, Director Office of NucleaI' Reactor Regulation U. F. NUCLEAR REG M TORY COMMISSION Washington, D. C. 20555 Attention: Mr. R. A. Clark, Chief s Operating Reactors Branch 3 Gentlemen:

DOCKET NO. 50-266 STEAM GENERATOR REPLACEMENT POINT BEACH NUCLEAR PLANT, UNIT 1 This letter provides documentation of the telephone conversation between Messrs. Colburn and Spraul of the NRC staff (NRC) and Messrs. Frieling, Seizert, Krieser, and Stevens of Wisconsin Electric Power Company (Licensee) on December 15, 1982. The purpose of the telephone conversation was to provide

~

clarification to Licensee's letter dated October 27, 1982 regarding commitments to regulatory guides and the quality assurance program as applicable to the Unit 1 steam generator replacement project.

The regulatory guide commitments by project. phase are as follows:

. 1. Fabrication The regulatory guides which are applicable to the fabrication phase are outlined in Section 2.1.4 of the O, Point Beach Nuclear Plant Unit 1 Steam Generator Repair Report (Repair Report). In addition, commitment was made to Regulatory Guides 1.58, 1.88, 1.144, and 1.146 in our October 27 letter.

2. Design O

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The regulatory guides which are applicable to the design phase are addressed in Section 3.6.3 of the Repair

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Report by reference to the. governing Westinghouse document, WCAP-8370, Revision 9A, Amendment 1.

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. Mr. H. R. Denton December 22, 1982 i 3. Installation The regulatory guides'which are applicable to the installation phase are addressed in Section 3.6.1 of the Repair Report by reference to Appendix H (currently section 1.8) of the Point Beach Nuclear O Plant Final Safety Analysis Report. In addition, commitment was made to Regulatory Guides 1.39, 1.58, 1.88,,1.94, 1.144, and l'.146 in our october 27 letter.

The reference to Sections 3.6.3 and 3.6.4 of the Repair Report in our October 27 letter as being applicable to installation was in error.

<O To provide further clarification, regulatory guide

" alternatives" as attached to our October 27 letter, pertain to the Westinghouse implementation of each respective regulatory guide and are not applicable to the Wisconsin Electric Quality Assurance Program.

Revision of the Repair Report will be made to incorporate ,

this and other additional information requests. Also, as advised during the referenced telephone conversation, NRC review of the supplement to Westinghouse WCAP-9245 is not necessary. Thus, the commitment in our October 27 letter is no longer required.

l' Please notify us if additional clarification is needed in this regard.

! Very truly yours, Assistant Vice President C. W. Fay Copy to NRC Resident Inspector O

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l%SC00 Sin Elecinc rowencoww O- 231 WEST MICHIGAN, MILWAUKEE. WISCONSIN 53201 ,

October 27, 1982 Mr.' H. R. Denton, Director ,

Office of Nuclear Reactor Regulation ,

U. S. NUCLEAR REGULATORY ComISSION Washington, D.C. 20555 Attention: Mr. Robert A. Clark, Chief O

• Operating Reactors, Branch 3 Gentlemen:

DOCKET 50-266 REQUEST FOR ADDITIONAL INFORMATION POINT BEACH UNIT 1 STEAM GENERATOR REPLACEMENT This is to provide further information, as requested by Mr. T. Colburn of your staff, regarding Quality Assurance (QA) programs described in the Point Beach Nuclear Plant Unit 1 Steam Generator Repair Report (" Repair Report"). As l

a matter of clarification, the Repair Report Section 2 describes the design and fabrication of replacement steawi generators and Section 3 describes the on-site l f,T V installation of the steam generators. The Regulatory Guides which are listed in Section 2.1.4 of the Repair Report are those which are applicable to design and l

fabrication. Those which are applicable to the installation are listed in Sections 3.6.3 and 3.6.4 of the Repair Jteport.

j Mr. Colburn requested enemitments to Regulatory Guides 1.39, 1.58, 1.88, 1.94, 1.144, and 1.1#,6 for the stesa generator replacement activities described in the Repair r : ort. The Repair Report will be revised to include commitments to the guid wce of Regulatory Guides 1.58, 1.88, 1.144, and 1.146 as

~

applicable to the design and fabrir.ation of the replacement steam generators and Regulatory Guides 1.39, 1.58, 1.88, 1.94, 1.144, and 1.146 as applicable to the installation of the str...a generators with the cir:~ifications as noted in Attach-IO ment 1 to this letter. Clarifications of Regulatory Guides applicable to the design and fabrication of the replacement steam generators are also contained in WCAP-8370. Reference to WCAP-8370 will be made in Section 3.6.4 of the Repair Report, ai: requested by Mr. Colburn.

Mr. Colburn also rsquested submittal of WCAP-9245 for review. Westing-house Electric Corporation is currently preparing a supplement to the WCAP-9245 O (Revision 6). This suppleoent will specifically address steam generator replace-ment services. Wisconsin Electric will review and approve the Westinghouse pro-gram, including tha supplc sent when it becomes available, in accordance with the l Wisconsin Electric Quclity Assurance Program, as indicated in Section 3.6.1, l paragraph 2, of the Repair Report.

! N

s Mr. H. R. Denton October 27, 1982 WCAP-9245 (Revision 6) is attached as Attachment 2 for your information and use in the review o.f the QA programs as applicable to the Point Beach steam p- generator replacement activities. The supplemental document to WCAP-9245

- (Revision 6) will be submitted for your information when it becomes available.

Should you havs further questions regarding QA programs to be imple-mented for replacement of steam generators at Point Beach Nuclear Plant Unit 1, please contact us.

Very truly yours, a

C. W. Fay Y' l c-i.A nistant Vice President Attachments cc (w/ Attachment 1): ASLB Service List NRC Resident Inspector bec: (w/ Attachment 1): R. W. Britt Sol Burttein R. H. Gorske/A. W. Finke D. K. Porter J. J. Zach D. B. Tate D.11. Stevens Gerald Charnoff

J. Woeber (W) l B. Spezialetti (W)

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WR W Powin counNr 231 W. MicillGUI, P.O. BOK 2044, MILWAUKEE, WI s3201 December 23, 1982 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U.S. NUCLEAR REGULATORY COWIISSION Washington, D.C. 20555 -

- Attention: Mr. R. A. Clark, Chief Operating Reactors Branch 3 Gentlemen:

DOCKET NO. 50-266 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION POINT BEACH NUCLEAR PLANT UNIT 1 This is in response to your request for additional information regard-ing Point Beach Nuclear Plant Unit I steam generator r'epairs transmitted by Mr. R. A. Clark's letter of November 16, 1982. Our responses to these requests O are attached.

In addition to the attached responses, we reviewed a request by tele-phone from Mr. Colburn of your staff for the approximate number of personnel who could be expected to receive radiation exposure during the steam generator repair and related activities. This is to confirm our estimate at that time of approximately 900 personnel who could receive radiation exposure during activ-ities associated with steam generator repair.

Should you have further questions, please contact me.

Very truly yours, at C. W. Fay Ass [stantVice sident Attachment I cc: ASLB Service List NRC Resident Inspector Blind copies to Messrs. R. W. Britt, Sol Burstein, R. H. Gorske/

A. W. Finke, D. K. Porter >, J. J. Zach, Os Gerald Charnoff

LICENSEE'S RESPONSE TO NRC REQUEST FOR

.'" ADDITIONAL INFORMATION RELATED TO POINT BEACH NUCLEAR PLANT UNIT 1 STEAM GENERATOR REPAIR d QUESTION 1 Describe stress relief heat treatment procedures of welded joints to ensure compliance with ASME Code requirements. Also indicate how the stresses would be l~ minimized on cladded components during cutting, welding and stress relief heat j , treatment.

RESPONSE

Concerning stress relief heat treatment for field installation, Westinghouse  ;

will comply with the requirements of the ASME Code Section III, Class I, N8-4620, 1977 Edition through the Summer 79 Addenda as applicable for each type of weld.

l Stress relief heat treatment industry practice, typical of that used at the Turkey I- Point and Surry plants during steam generator repairs, will be used for the Point Beach Unit 1 steam generator repair. Special work packages for these activities will be used during the steam generator repair. .

O With regard to welding of cladded components, it should be noted that the Point Beach steam generator repair methodology does not involve any field welded joints ,

between cladded components. The steam generator channel head cladding is applied at the manufacturing facility. Safe ends are in:talled on the inlet and outlet reactor coolant nozzles at the manufacturing facility to facilitate field welding of the nozzles to the stainless steel pipe fittings without the need for stress relief heat treatment. Therefore, there will be no welding, cutting, or stress relief heat treatment effects of cladded components.

QUESTION 2 Describe the details of the tests and evaluations which will be conducted after the steam generator repair to assure the integrity of the reactor coolant system and compliance with applicable codes and regulations.

O

~ RESPONSE As described in Section 3.2.6.7 of the Repair Report, hydrostatic tests and base-line inspections of piping or components will be performed in accordance with Section XI of the ASME Code for piping or components affected by the steam gen-crator repair. These hydrostatic tests and inspactions will assure the integrity cf the reactor coolant system.

In addition to the hydrostatic tests following repair, the replacement steam generators will be hydrostatically tested on the A baseline eddy current inspection primary side at the manufacturing facility.

of 100 percent of the steam generator tubes will be conducted following comple-tion of the repairs.

Detailed procedures have not born developed for these hydrostatic tests and in-spections at the present time.

QUESTION 3 i

' O Describe the preoperational testing program which will be co can be operated in accordance with design requirements and in a manner that will not endanger the health and safety of the public.

RESPONSE

As described in the Repair Report, the repair of steam generators in Unit I will have minimal impact on existing equipment and structures except for l

l the steam generators and associated piping and instrumentation in containment.

Therefore preoperational testing requirements will be similar to those for a normal refueling outage. .

O O l

Up:n completten of the steam generater repair, piping systems will be hydro-statically tested in accordance with applicable codes. Instrumentation and elec-trical equipment which was removed or relocated during the steam generator repair w111 be reinstalled and tested using normal maintenance procedures and Technical Specification requirements, as necessary. Other systems and components will be tested in accordance with Technical Specification requirements for testing prior toreturntopowerfromanormalrefueling.

Testing of'the repaired steam generators and other affected systems or components will include the following:

1. Steam generator thermal performance tests to verify the thermal perform-ance parameters specified for the repaired steam generators.

2. Steam generator moisture carryover tests to verify that moisture carry-over in the steam is within design values.

O 3. Calorimetric tests to verify adequate reactor coolant flow in accord-ance with Technical Specifications.

4. Inspection of equipment supports affected by repair activities in both hot and cold conditions.

These tests and inspections are not significantly different from_ tests and inspections performed following normal refueling or maintenance and provide adequate assurance that operation of the repaired steam generators will not affect the health and safety of the public.

i QUESTION 4 O .

As referenced in Section 3.2.6.2 of the Steam Generator Repair Report, describe the areas / components that will be decontaminated and subsequently placed back in service. Describe the decontamination process including the decontamination fluid. Describe the tests that have been performed to show that decontamination O fluids are benign and will not cause future corrosion.

RESPONSE

Decontamination, as generally referenced in Section 3.2.6.2, consists of removal O of radioactive contamination, to the extent practical, from work areas of the containment frequented by personnel during the steam generator repair. The

, decontamination of these areas will be accomplished using techniques the same as, or similar to, those in use during normal plant operation and maintenance.

Such techniques include flushing of surfaces with water, mopping, wiping with absorbent cloth or covering of floor surfaces. As stated in Section 3.2.6.4, local areas of containment will be decontaminated to minimize personnel exposure to radioactive contamination. As noted in Section 3.3.5.3, local areas of the reactor coolant piping and steam generator surfaces may be decontaminated if nec-essary and local work areas will be decontaminated periodically. These activities do not result in introduction of decontamination materials to systems or components affected by the steam generator repair.

As stated in Table 5-1, Item 3.(4) of the Repair Report, and our November 22, 1982 response to NRC Staff requests for additional information, no chemical decontami-1 nation of components is planned for the repair process. Thus, chemical decontami-nation fluids will not be introduced to systems or components affected by the repair. Reactor coolant pipe ends will be wiped with lint free rags dampened with domineralized water to reduce surface contamination. While details of weld preparation procedures have not been finalized, additional decontamination may be performed using decontamination methods with abrasive material such as boric acid grit in a water slurry if determined to be required based upon AL4RA eval-untion. Such decontamination would not introduce meterials deleterious to the reactor coolant system.

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! QUESTION 5 As referenced in Section 3.2.6.4 of the repair report relating to post-shutdown activities, describe the water chemistry and/or metal surface conditions to be established and maintained in parts of primary and secondary systems not being refurbished or replaced.

RESPONSE Layup proceduras for primary and secondary systems during steam. generator repair are presently being developed. It is expected that wet layup using borated, de-eineralized water will be used for primary systems with hydrazine addition and/or nitrogen blanketing, as appropriate, to prevent intrusion of oxygen. Layup of secondary systems such as the feedwater system and condenser, is expected to be dry. These systems will be drained and dried with air in accordance with normal practice.

Detailed layup procedures have not been developed presently.

O QUESTION 6 Verify the adequacy of the spent fuel pool water cooling and clean-up systems to handle the off-loading of the full core.

RESPONSE

The Point Beach Nuclear Plant Final Safety Analysis Report (FSAR) con-tains a description of the spent fuel pool cooling and purification systems and design bases for the cooling system. As described in Section 9.3.1 of the FSAR, the design basis for the spent fuel pool cooling system assumes a full core unload for inservice inspection or maintenance of reactor coolant system compon-ents with spent fuel assemblies stored in all other spent fuel rack storage locations. Thus, the spent fuel pool cooling system is adequate to accommodate the core unload required for repair of the Unit 1 steam generators.

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The core unload required for steam generator repair would not be ex-pected to result in radioactivity concentrations significantly diffurent from

.those resulting from storage of spent fuel, refueling operations or unloading of the core for routine inservice inspection. Experience during past core unloads and nomal refuelings at Point Beach has indicated that the spent fuel pool purification system is adequate to maintain spent fuel pool radioactivity concen-trations at acceptably low levels. Thus, the spent fuel pool purification system is adequate for the core unload required for steam generator repair.

QUESTION 7 As described in Section 3.2.6.7 of the repair report relating to post-installation activities, provide a commitment to inspect, after hydrotest, the interior of the steam generator to assure metal-clean surfaces in accordar.ca with Regulatory Guide 1.37.

RESPONSE

l Cleaning criteria during fabrication and preparation for shipment of the steam g~erators are specified in WCAP-8370, Rev. 9A which includes criteria for both the primary and secondary sides of the stora generators. The steam generators are sealed and protected against moistura during shipment and are maintained dry during installation. Thus, there is assurance that the steam generator surfaces i will not be exposed to wet, oxygenated environments during installation and that primary and secondary surfaces are acceptably clean. Reopening the steam gener-ator for an inspection following hydrostatic testing would expose surfaces to a wet, oxygenated environment which could result in formation of iron oxides on

carbon steel steam generator surfaces in the secondary side. As a consequence, the " metal clean" criterion implied by ANSI N45.2.1-1973 paragraph 3.1.2(1) (as referenced in Regulatory Guide 1.37) probably could not be met on the secondary side of the steam generators following hydrostatic testing and cannot be met following a period of operation since the carbon steel surfaces would be covered

f with the normal filo of magnetite cnd other iron oxides. Since primary surfcc::s are41therInconelorStainlessSteelandareverifiedcleanfollowinghydrotest at the manufacturing facility, there is no basis for requiring an inspection following hydrotesting in the plant.

QUESTION 8 Section 3.4.2 of the repair report relates to on-site storage of components.

Provida the details of the lower assembly sealing prior to storage. Address the thickness of the seal plates and welds and the preparation of the interior of the assembly, i.e., drying, gas cover, etc.

j RESPONSE The removal procedures require that the steam generator lower assembly be seeled j prior to movement out of the containment. Seeling will be accomplished by welding I

closure plates (covers) over the top of the lower assembly at the girth cut loca-

> tion and over the inlet and outlet reactor coolant nozzles and all other vessel

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penetrations. The sealing is accomplished to assurir containment of the radio-activity within the lower assembly and to minimize the potential for radiation streaming from these penetrations. Thickness of the covers for the top of the lower assembly and nozzles will be such that structural integrity is assured

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and that radiation levels are reduced to acceptable levels. Estimated thick-I ness of the covers is approximately three (3) inches of steel or its shielding equivalent. The steam generator lower assembly primary and secondary sides will be drained. Essentially all the radioactivity is contained in the primary, side of the steam generators which will be dry during the steam generator repair and l

l storage. Thus, there is no need for further drying or gas blanketing of the steam generators..

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