ML20062K982

Steam Generator Repair Rept

ML20062K982
Person / Time
Site:	Point Beach
Issue date:	08/31/1982
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20062K948	List: ML20062K946 ML20062K961 ML20062K982
References
NUDOCS 8208180086
Download: ML20062K982 (185)

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WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT l UNIT NO. 1~

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STEAM GENERATOR REPAIR REPORT 1

AUGUST 1982 l

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TABLE OF CONTENTS O

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SUMMARY

AND CONCLUSIONS

c 1.1 Sunaary of Steam Generator Repair Program 1-1 1.1.0 ' Introduction 1-1 1.1.1 Contairnent Entry and Exit of Steam Generator Lower Assemblies 1-2 m 1.1.2 Steam Generator Lower Assembly Characteri stic s 1-2 1.1.3 Safety Related Considerations 1-2 1.1.4 ALARA Considerations 1-3 l

l 1.1.5 Off site Radiological Considerations 1-3 1.1.6 Other Aspects of the Program 1-3  !

l.1.7 Steam Generator Di sposal 1-4 1.2 Identification of Principal Agents and Contractors 1-4 1.3 Other Considerations 1-4 1.4 Conclusions 1-5 l 2.0 REPLACEMENT COMPONENT DESIGH ,

2.1 Compcri son with Existing Component 2-1 2.1.1 Parametric Comparison 2-1 ,

2.1.2 Physical Coupatibility with Existing -

Steam Generators and Systems 2-2 l

2.1.3 ASME Code Applications 2-2 2.1.4 Regulatory Guide Application 2-2 2-6 1 'Aq 2.2 Componerit Design Improvements 2.2.1 Design Refinement: to Minimize the Potential for Corrosion 2-7 2.2.2 Design Refinements to Improve Performance 2-10 V 2.2.3 Design Features to Permit Ease of Maintenance and Reliability 2-11  !

l l 2.3 Shop Tests and Inspections 2-12 lO ,

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, TABLE OF CONTENTS (Continued)

Page No.

L j 3.0 COMPONENT REPLACEENT PROGRAM AND PROCEDURES 3-1 j 3.1 Pathways and Construction Restr'ictions 3-2 l- 3.1.1 Site Preparation 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 3.2 Equipment and Concrete Removal and Replacement 3-7 3.2.1 Mechanical Equipment 3-8 3.2.2 Instr:1 mentation 3-8 3.2.3 Cable and Conduit 3-8 3.2.4 Piping 3-9 3.2.5' Concrete and Structural Steel 3-9 3.2.6 Removal and Installation of Steam O

V 3.3 Generators Radiological Protection Program 3-10 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 Requirements 3-24 3.3.5 General ALMA Considerations 3-26 3~.3.6 Miscellaneous Waste Disposal 3-29 v 3.4 ' Disposition of Steam Generator Lower Assemblies 3-31 7 3.4.1 Objectives of Handling / Disposal Operations 3 - 31 F 3.4.2 Onsite Storage 3-31  :

3.4.3 Offsite Disposal 3-32 O~ 3.4.4 Radioactive Releases and Dose Assessment Associated with Onsite Storage 3-32  ;

3.4.5 Accident Considerations Associated with Onsite Storage 3-33 I

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TABLE OF CONTENTS (Continued)

Page No.

3.4.6 Conclusions 3-35 3.4.7 References 3-35 <

3.5 Plant Security 3-36 3.6 Quality Assurance Program 3-36 3.6.1 WE Quality Assurance Program 3-36 i 3.6.2 Westinghouse Nuclear Service Division (WNSD) Quality Assurance Program 3-36 3.6.3 Westinghouse Nuclear Technology Division (WNTD) Quality Assurance Program 3-37 3.6.4 Westinghouse Nuclear Components Division (WNCD) Quality Assurance Program 3-37 4.0 RETURN-TO-SERVICE TESTING 4-1 8 5.0 SAFETY EVALUATIONS 5.1 FSAR Evaluations 5.1.1 Introduction 5-1 5-1 5.1.2 Non-LOCA Accidents 5-1 5.1.3 Loss-of-Coolant-Accident (LOCA) Evaluation 5-16 5.1.4 Steam Generator Tube Rupture 5-19 5.1.5 References 5-19 5.2 Construction Related Evaluations 5-21 5.2.1 Handling of Heavy Objects 5-21

( 5.2.2 Radioactive Release and Dose Assessment 5-25 5.2.3 References for Section 5.2 5-29 6.0 ALARA Considerations y/ 6.1 ALARA Objectives 6-1 l

6.2 Steam Generator ALARA Program 6-2 l 6.3 Training and Instruction 6-4 6.4 Engineering and Design Reviews 6-5 l v' 6.5 Design Features 6-6 iii 2671Q: 1

TABLE OF CONTENTS (Continued) l Page No. l 6.6 Radiological Impact 6-8 6.6.1 In-Plant Doses 6-8 6.6.2 Doses to the Public 6-9 7.0 Environmental Aspects of the Repair 7.1 General 7-1  ;

O' 7.2 Resources Committed 7-1 7-1 7.2.1 Non-Recyclable Building Materials 7.2.2 Land Resources 7-1 7.2.3 Water Resources 7-3 7.3 Waste Water 7-4  ;

7.3.1 Sanitary Facilities 7-4 7.3.2 Laundering Operations 7-5 7.4 Construction 7-5 8 7.4.1 Noi se 7.4.2 Dust 7.4.3 Open Burning '

7-5 7-5 7-6 7.5 Radiological Aspects 7-6 7.6 Return of Operation 7-6 7.6.1 Water Use 7-6 7.6.2 Operational Exposure 7-7 7.6.3 Radiological Releases 7-7 O  ;

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LIST OF TABLES O Table No. Title 2-1 Steam Generator Design Data (Per Steam Generator)

O 3-1 Typical Portable Survey Instrument Specifications 3-2 Comparison of Estimated Occupational Doses for Steam Generator Disposal Alternatives (Man-Rem) 5-1 Comparison of Operating Parameters for Original and Repaired Steam Generators 5-2 Comparison of Design Parameters for Original and Repaired Steam Generators 5-3 Gross Contamination Levels by Location in Piping and Steam

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Generator 5-4 Point Beach Nuclear Plant, Unit 1 Estimated Steam Generator Curie Content 5-5 Effluent Release Isotopic Distributions Steam Generator Replacement Project Surry Power Station - Unit No. 2 5-6 Comparison of Gaseous Effluent Releases O 5-7 Radionuclide Concentrations in Reactor Coolant 5-8 Estimated Specific Activities of Laundry Waste Water 5-9 Estimated Radionuclide Releases Due to Discharge of Reactor Coolar.t Water O

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i LIST OF TABLES (Continued)

Table No. Title

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5-10 Estimated Radioactive Liquid Effluent Releases During the i Steam Generator Repair f

l 5-11 Comparison of Radioactive Liquid Effluent Releases f

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6-1 ALARA Provisions

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'i 6-2 Estimate of Personnel Radiation Exposures for Steam l Generator Replacement Operations at Point Beach Unit 1 l

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l LIST OF FIGURES <

. I O Figum No. Title  !

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Figure 2-1 Steam Generator Lower Assembly .

t O Figure 2-2 Flow Distribution Baffle and Blowdown i

l Figure 2-3 Quatrefoil Tube Support Plate Schematic  !

.i O Figum 2-4 Tube to Tubesheet Junctum t

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Figure 3-1 Outage Sequence i ) l Figure 3-2 Lower Assembly Removal Sequence (Sheet 1 of 2) f i

Figum 3-2 Lower Assembly Removal Sequence (Sheet 2 of 2)

Figure 3-3 Steam Generator Laydown Areas O Figure 3-4 Steam Generator Haul Routes l

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Figure 3-5 Reactor Coolant Piping Cut Points i t

Figure 3-6 Feedwater and Steam Line Piping Cut Points [

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SUMMARY

AND CONCLUSIONS d 1.1

SUMMARY

OF STEAM GENERATOR REPAIR PROGRAM t

1.1.0 Introduction O Point Beach Nuclear Plant Unit I has experienced secondary side corro-sion in a number of tubes in the two steam generators. Various amelio-rative measures have been taken to arrest the corrosion, including changes in the secondary water chemistry, plugging degraded tubes, and Ox reduction of operating temperature. Approximately 14 percent of the tubes in each steam generator have been removed from service by plug-ging. As a result of the reduced operating temperature, Unit 1 is cur-rently operating at less than 80 percent of full power. To increase availability and reliability, and to return to full-power operation, it is appropriate to replace both steam generators of Unit 1.

This document discusses the steam generator repair program which will be j

. implemented to restore the reliability and perfonnance of the steam  ;

( generators installed in Unit 1. The discussion of the steam generator .

repair program and the effect on the operating unit demonstrates that the repair work and subsequent operation can be conducted without undue l risk to the health and safety of the general public or to personnel engaged in the repair work. The information contained herein is not intended to supplant the information in the Final Safety Analysis Report (FSAR), but is intended to supplement the discussion presented tnerein in the specific areas associated with the repair program and to identify [

i significant changes that may result from the repair program. The FSAR  !

for the Point Beach Nuclear Plant snould be consulted for specific l details about referenced equipment, systems, or components.

l t The information presented herein reflects the most current design infor-mation at the time of preparation. Since the detail design and engi-neering for the program are currently in progress, this document will be

! revised as new information is developed.

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1.1.1 Containment Entry and Exit of Steam Generator Lower Assemolies O '

V Entry and exit of the steam generator lower assemblies will be through the present equipment hatch in the containment structure. The equipment hatch is sized to accommodate steam generator replacement witnout con-tainment modification. The steam generator lower assemblies will be moved tnrougn the equipment natch using a temporary containment trans-port system.

1.1.2 Steam Generator Lower Assembly Characteristics O ,

Westinghouse Electric Corporation will fabricate new steam generator lower assemblies. The design of the lower assemolies will match the design performance of the lower assemblies being replaced. tiowever, several design features that do not alter mechanical performance and FSAR parameters are included in the design. These design features will provide improved thermal hydraulic performance, improved access to the -

tube bundle, and reduce the potential for secondary side corrosion.

1.l.3 Safety-Related Considerations i

The potential impact of the repaired steam generators on each appro-  :

priate accident analyzed in the FSAR has been evaluated. In addition, f:

it is realized that the repair effort involves extensive work with l radioactive components which include cutting, welding and transporting of portions of the steam generators and associated piping. Point Beach Nuclear Plant has had extensive experience througnout its history in l similar activities. Because of the essentially duplicated safety-l related design parameters; improved thermal-hydraulic, corrosion resis-l tance, and maintainability characteristics; and previous experience, it is concluded that tne repair work and subsequent operation can be con- '

ducted without undue risk to the health and safety of the general public  :

or to personnel engaged in the repair work and does not involve an unre- i viewed safety question. 1 i

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1.1.4 ALARA Considerations O The guidelines contained in Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposure at Nuclear Power Sta-tions Will Be As Low As Is Reasonably Achievable," will be considered.

The entire repair process will be preplanned. Mockups and training will O' be used extensively to minimize outage time and radiation exposure.

Decontamination and other exposure limiting techniques will be used where they offer significant savings in exposure commensurate with over- i all program objectives. Special scaffolding and other components will O,

be prefabricated to the extent possible to minimize radiation exposure and outage time.

In addition, estimates have been made of the exposure to personnel i involved in the repair activity. This evaluation indicates that the reduction in man-rem exposure currently being incurred during tube r inspection and plugging operations will offset, in a short time period, the man-rem exposure incurred during the steam generator repair.

1.1.5 Offsite Radiological Considerations Radiological controls will be in effect during all activities associated with the repair program. Since the lower steam generators may be a source of radioactive contamination during and following cutting opera-tions, special covering devices will be employed to minimize radiation exposure and the spread of radioactive contamination.

I 1.1.6 Other Aspects of the Program The shop fabrication of the lower assemblies and moisture separator replacement parts will be conducted in accordance with standard prac-1 tices. Transport, lifts, removal and replacement of components, and site preparation associated with the repair program will utilize stan-dard manufacturing and construction practices.

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1.1.7 Steam Generator Disposal The repair activity and ultimate disposal of the existing lower assem-blies are separable issues. During the time between removal from con-tainment and ultimate disposal, the lower assen.Olles will be stored q onsite in a temporary storage facility.

b 1.2 IDENTIFICATION OF PRINCIPAL AGENTS AND CONTRACTORS The Wisconsin Electric Power Company (WE), is the sole owner and O licensed operator of Point Beach Nuclear Plant. WE has been actively engaged in nuclear power operations with the construction, operation, and maintenance of Point Beach Nuclear Plant Units 1 and 2. This repre-sents a total operating experience of.approximately 22 reactor years. j Westinghouse Electric Corporation (Westinghouse) manufactured the exist-ing steam generators and designed and fabricated the replacement steam generator lower assemblies and moisture separator components. Westing-house experience in nuclear plants for the electric utility industry is demonstrated by the pressurized water reactor planb that Westinghouse has designed, developed, and manufactured. Westinghouse has developed a broad technological foundation in nuclear power application which enables them to offer the electric utility industry a reliable and safe-source of power and services related to the maintenance of nuclear power plants.

i 1.3 OTHER CONSIDERATIONS l

The repair program will involve replacement / repair of facility equip-

ment, rather than an alteration or change to the f acility. Because of the scope of the steam generator repair, tne process has been reviewed by WE pursuant to 10 CFR 50.59. In addition to eacn FSAR accident analysis evaluation, tne construction incident potential, the potential impact on the ability to shut down the operating unit and maintain it in a safe shutdown configuration, and the impact on cooling spent fuel have O l r

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also been evaluated. The evaluations indicate tnat the repair activity f does not involve an unreviewed safety question, a change to Point Beacn O Nuclear Plant Unit 1 Tecnnical Specifications is not required, and tne

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repair work and subsequent operation can be conducted without undue risk l to the health and safety of the general public or to the personnel f

engaged in the repair work.

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1.4 CONCLUSION

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The fundamental conclusions reached are that the steam generator repair j O program can be conducted utilizing proven manufacturing and construction j techniques and that the repair program does not result in any adverse j impact on plant safety. The repair effort will provide employment, (

income, and sales revenue to the local region and will not significantly affect the environment of the plant site or immediate adjacent areas. {

The detailed bases supporting these conclusions are provided in tne  :

report tnat follows.

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2.0 REPLACEENT COMP 0NENT DESIGN Westinghouse will shop fabricate new steam generator lower assemblies as my illustrated by Figure 2-1. Tne design of the lower assemblies will be similar to the design performance of the lower assemblies being replaced. However, several design features which do not alter mechan-ical, performance and Final and Safety Analysis Report (FSAR) parameters V are included in the design. These design features will provide improved flow distribution, access improved to the tube bundle, and will reduce tne potential for secondary side corrosion. This section discusses the A design and manufacture of the lower assemblies.

2.1 COMPARIS0N WITH ORIGINAL COMPONENT 2.1.1 PARAETRIC COMPARISON The steam generators for the Point Beach Nuclear Plant plant Unit 1, upon completion of the repair, will have physical, mechanical and ther-mal characteristics consistent with the original design and safety 8 analysis as currently documented in the FSAR.

Design data for the existing and repaired steam generators are presented i n Table 2-1. The thermal performance data for each steam generator will remain the same as the original steam generators.

Materials used in the fabrication of the replacement lower assemblies will be identical to those used in the original steam generators except W1ere specific design changes have been incorporated or fabrication v' practice has changed. These changes include the following:

1. Plate material used in the secondary shell formation has been changed to SA-533 Grade A Class 2 from SA-302 Grade B Class 1 as a O

v result of changes in fabrication practices; s

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2. Support plate material has been changed to SAS-240 Typ2 405 from SA-285 Grade C to minimize corrosion and the potential for denting; and v

3. The steam generator tube material for the replacement steam genera-tor assemblies is thermally-treated Inconel 600. The original tube

/N 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-ther discussion of mateial changes is provided in Section 2.2 Table 2-2 provides comparison of past and present aplications of mateials.

2.1.2 PHYSICAL COMPATIBILITY WITH ORIGINAL STEAM GENERATORS AND SYSTEMS The replacement steam generator lower assemblies are designed to be duplicate physical replacements for the existing units. Outside overall dimensions are the same as are the locations of .zles and support attachments. Existing interfaces between the steam generators and plant 8 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.

2.1.3 ASE CODE APPLICATION l

The original steam generators were built to the 1965 Edition of the ASE Boiler and Pressure Vessel Code (ASE Code), including Addenda through Summer 1966; the replacement steam generator lower assemblies will be

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

V 2.1.4 REGULATORY GUIDE APPLICATION The compilation below addresses Regulatory Guides considered applicable to the fabrication of the replacement lowr assemblies. It must be noted that these guides were issued subsequent to construction and 2284Q: 1 2-2

operation of this facility. The intent is to accomodate, consistent with facility design and repair program objectives, the guidance pro-vided by these regulatory guidelines.

V 1.26 Quality Group Classifications and Standards for Water, Steam and Radioactive-Waste-Containing Components of Nuclear Power Plants (Rev. 2) July 1975.

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1.28 Quality Assurance Program Requirements (Design and Construction)

(Rev. 2) Feb.1979 (Safety Guide 28, June 1972) d The Westinghouse position on Regulatory Guide 1.28 is presented in WCAP-8370, Revision 9A, "WRD Quality Assurance Plan." i 1.29 Seismic Design Classification (Rev. 3, July 1978).

1.31 Control of Stainless Steel Welding (Rev. 3) October 1978 The Westinghouse production weld verification program, as described in WCAP-8324-A, was approved by the NRC as a satisfac-tory substitute for following the recommendations of the NRC Interim Position on Regulatory Guide 1.31 (4/74). The results of the verification program support the hypothesis presented in WCAP-8324-A; these results have been summarized and documented in WCAP-8693, which has been submitted to the NRC for information.

1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (Sept.,

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i 1981) 1 The Westinghouse position on Regulatory Guide 1.37 is presented in O) qi WCAP-8370, Revision 9A, "WRD Quality Assurance Plan."

1.38 Qulity Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling for Nuclear Powr Plants (Rev. 2, May 1977) i t

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The Westinghouse position on Regulatory Guide 1.38 is presented in WCAP-8370, Revision 9A, "WRD Quality Assurance Plant."

1.43 Control of Stainless Steel Weld Cladding of Low-Alloy Steel Compo-nents (May 1973)  ;

O The Westinghouse Nuclear Components Division will use materials made to fine-grain practice or which are not susceptible to under-clad cracking. These materials do not require the controls listed in the guide.

1.44 Control of the Use of Sensitized Stainless Steel (May 1973)

All of the unstabilized auctenitic stainless steels used for com-ponent parts of the reactor coolant pressure boundary are utilized in the final heat treated condition required by the respective ASE Code,Section II, material specification for the particular I type or grade of alloy. Processing and fabricction are performed using established methods and techniques to avoid sensitization.

O Westinghouse has verified that these practices will prevent sensi-tization by performing corrosion tests on as-received wrought l

material, as well as on production and qualification weldments.

In addition, the water chemistry in the reactor coolant system is l controlled to prevent intergranular attack of unstabilized stain- j less steels; the effectiveness of these controls has been demon- i I

strated by both laboratory tests and operating experience.

1.48 Design Limits and Loading Combinations for Seismic Category I Fluid Systems Components (May 1973) i Westinghouse meets and will continue to meet the requirements of General Design Criterion 2 and will thereby satisfy the concerns t of Regulatory Guide 1.48. The loading combinations and design limits used in the code stress analysis of the steam generator will be the same as those in the Point Beach FSAR.

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l.50 Control of Preheat Temperature for Welding of Low-Alloy Steel i (July,1976) i Westinghouse practices are in agreement with Regulatory Positions '

C.1.a. C.3 and C.4. For Regulatory Position C.1.b; Westinghouse j qualified Welding procedures within the preheat temperature ranges (

O required by Section IX of the ASME Code. For Regulatory Position C.2, Westinghouse uses the methods documented in WCAP-8577-A, i which has been accepted by the NRC.

I O 1.54 Quality Assurance Requirements for Protective Coatings Applied to [

Water-Cooled Nuclear Power Plants (Rev. 9, June 1973).  !

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1.60 Design Response Spectra for Seismic Design of Nuclear Power Plants (Rev.1, December 1973).

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1.61 Damping Values for Seismic Design of Nuclear Power Plants (Rev. 0, l October 1973). l 1.64 Quality Assurance position on Regulatory Guide 1.64 is presented in WCAP-8370, Revision 9A, "WRD Quality Assurance Plan." fF 1.68 Preoperational and Initial Startup Test Programs for Water Cooled Power Reactors (August,1978).

1.74 QA Tenns and Definitions (February,1974). l l

1.83 Inservice Inspection of Pressurized Water Reactor Steam Generator Tuf es (Rev.1, July 1975).

l 1.84 Code Case Acceptability-ASE Section III Design and Fabrication (Rev.19, April 1982).

O l.85 Code Case Acceptability-ASE Section III Materials (Rev.19, April 1982).  !

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1. Westinghouse controls its suppliers to:

a. Limit the use of code cases to those listed in Regulatory Position C.1 of the a.nplicable 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 A 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 NRC pe lission for the use of code ca.ses needed by suppliers and not yet endorsed in Regulatory Posi-tion C,1 of the applicable guide revision in effect at the O_ time the equipment is ordered and permits supplier use only if NRC pemission is obtained or is otherwise assurd (e.g., a later version of the regulatory guide includes endorsement). .

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

l 1.116 QA Requirements for Installation, Inspection, and Testing of ,

Mechanical Equipment and Systems (May 1977). l 1.121 'ases o 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).

2.2 COMPONENT DESIGN IMPROVEENTS  :

The physical, thermal 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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These features, will increase the operating reliability and reduce tha potential for corrosion of the steam generator components.

O O Extensive research, development and testing have been utilized to select design parameters, materials and component configurations diich will minimize the potential for corrosion and enhance the performance of the O

V repaired steam generators.

2.2.1 DESIGN REFINEENTS TO MINIMIZE THE POTENTIAL FOR CORROSION 2.2.1.1 FLOW DISTRIBUTION BAFFLE b[~N A flow distribution baffle has been provided 23 inches above the tube-sheet. This baffle has a cut out center section and oversized drilled tube holes. The baffle plate assists in directing flow across the tube-sheet then up the center of the bundle through the center cutout. The design is sized to maximize the flow to the center of the bundle and mini-mize the number of tubes in low-velocity regions. Consi stent with this purpose, the design is also intended to cause any sludge to deposit ne'ar 8 the blowdown intake where it can be removed. The flow distribution baffle plate material is ferritic stainless, steel. Figure 2-2 illustrates the flow distribution baffle. As noted in Section 2.2.3.1, access holes have been provided to allow sludge lancing above and below the baffle plate.

2.2.1.2 IMPROVED INTERNAL BLOWDOWN DESIGN Maintenance of the secondary side water chemistry is assisted through the use of the blowdown system. Each steam generator will be designed to have

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d two 2-inch schedule 40 Inconel internal blowdown pipes. The blowdown

! nozzles on the external portion of the steam generator shall have provi-sions for connection to 2-1/2 inch existing blowdown piping. The blowdown

! intake location is coordinated with the baffle plate design so that the maximum intake is located where the greatest amount of sludge may collect. The modified blowdown system should allow increased capacity blowdown in comparison with the present blowdown arrangement.

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2.2.1.3 TUBE EXPANSION IN TUBESHEET Following insertion into the tubesheet hole, tack rolling, welding and gas V leak testing, the tubes are hydraulically expanded to the full depth of the tubesheet holes. Full-depth closes eliminates the tube sheet crevice in which concentration of impurities has occured in the original steam p generator.

G 2.2.1.4 THERMALLY TREATED INCONEL 600 TUBING Research by Westinghouse has determined that additional resistance in the d 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 with grain boundary precipitate morphology, which provides increased mar-gin with respect to stress corrosion resistance. Several benefits results from this treatment such as additional resistance to stress corrosion cracking in NaOH, additional resistance to intergranular attack in oxygen-ated envirornents, additional resistance to intergranular attack in sul-8 phur-containing species and reduction of residual stress imparted by tube proce ssi ng. ,

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 v' 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 (O center of the bundle. This area is utilized for location of the blowdown D') i ntake. Feedwater flow distribution is accomplished by providing a grea-ter number of flow paths on the portion of the feedwater ring which traverses the hot leg side of the tube bundle.

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2.2.1.6 CORROSION RESISTANT SUPPORT Pl. ATE MATERIAL The support plate material has been selected such that the potential for J denting of the tubing due to corrosion in the crevice between the tube and tube support plate is significantly reduced. SA-240 Type 405 ferritic stainless steel has been selected for this application. Thi s material i s ASE Code-approved and is believed to be resistant to corrosion with the chemistry expected during the operation of the steam generator. In addition, SA-240 has a low wear coefficient when paired with Inconel and has a coefficient of thermal expansion similar to carbon steel. Corrosion 7 of SA-240 results in an oxide which has approximately the same volume as (d the parent material. In addition to the tube support plates, the baffle plate (discussed in Subsection 2.2.1.2) will be constructed of SA-240 Type 405.

2.2.1.7 QUATREF0Il TUBE SUPPORT PLATES The quatrefoil tube support plate design, illustrated by Figure 2-3, con-sists of four flow lobes and four support lands. The lands provide sup-8 port to the tube during all operating conditions, sile allowing flow around the tube. This design also djrects the flow along the tubes which limits steam formation and chemical concentration at the tube-to-tube support plate intersections. The quatrefoil support plate design results in higher average velocities along the tubes, which should sludge depo si tion. The combination of higher velocities in the support p1 ate region and corrosion resistant material will minimize the potential for support plate corrosion.

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("/ 2.2.2 DESIGN REFINEMENTS TO IMPROVE PERFORMANCE In the course of the steam generator design, as derived from operating experience and ongoing research and development programs, certain modifi-

\s cations and refinements have been incorporated in recent designs to provide additional performance of thermal hydraulic characteristics.

These are included in the Point Beach design and are discussed below.

They do not alter previous safety analyses.

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2.2.2.1 FLUSH TUBE TO TUBESHEET WELD p The tubes on the replacement lower assemblies will be flush with the tube-D sheet holes and then welded to the tubesheet cladding. Elimination of the protruding tube stub of the original design results in lower entry pres-sure losses and, therefore, a lower pressure drop in the primary loop. In addition, a possible point of crud buildup and corrosion is minimized with thi s design. This illustrated in Figure 2-4.

2.2.2.2 TUBE LANE BLOCKING DEVICE b

b A portion of the recirculated water exiting at the bottom of the wrapper will tend to preferentially channel to the tube lane and bypass part of the tube array. In order to minimize this tube bundle bypass, a series of plates are installed in the tube lane to block the bypass flow paths.

These plates are compatible with sludge lancing.

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

be replaced with a primary moisture separator assembly consisting of one hundred and twelve modular 7" I.D. swirl vane assemblies. These modifica-tions provide improved steam-water separation and reduced moisture carryover.

2.2.3 DESIGN FEATURES TO PERMIT EASE OF MAINTENANCE AND RELIABILITY O

\w Operational experience, including necessary maintenance and repair, has led to certain changes in design with the objectives of increasing providing additional maintainability of the units. These changes are dis-cussed below and do not affect performance or FSAR safety analyses.

v 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-2284Q: 1 2-10

sheet, approxihlatoly 90 dagrees apart, with tw located on tha tube lane-below the baffle plate. Two 6-inch access ports will be located on the tube lane, between the baffle plate and the first tube support plate. The addition of these access ports should promote inspection of the tubesheet and flow distribution baffle plate .

2.2.3.2 INSPECTION PORT One 3 inch inspection port is located on the lower shell transition cone at an elevation slightly above the top tube support plate of the tube O

G bundle.. This port is located on the tube lane centerline and provides for inspection of the top support plate and the tubing U-bend area.

2.2.3.3 WET LAYUP N0ZZLE A 2-inch nozzle is to be added to the upper shell to facilitate the wet layup of the steam generators during periods of inactivity. The wet layup nozzle can be used during these periods to maintain desired water chemis-try in the steam generator. The nozzle can also be used in conjunction 8 with other equipment to circulate water through the steam generator during periods of layup. ,

2.2.3.4 PRIMARY SHELL DRAIN A 3/8-inch primary shell drain is included in the channel head to provide additional drainage of the channel head. This drain facilitates mainte-nance and inspection to be conducted in the channel head.

2.2.3.5 PRIMARY CLOSURE RINGS l

! Closure rings will be welded inside the channel head at the base of each primary nozzle so that closure plates can be installed during primary x " ,)

chamber maintenance. This design allows the plates to be bolted to the rings for quick installation and removal. Closure plates allow mainte-nance or inspection to be conducted in the channel head with the reactor j cavity flooded and, thhhus, can reduce outage time.

2284Q: 1 2-11

2.2.3.6 STEAM N0ZZLE FLOW LIMITING DEVICE A flow limiting device will be provided to be installed in the steam out-G's let nozzle to minimize the pressure drop across internal components during a postulated 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 conducted during the fabrication of the steam generator lower assembly.

In addition to these ASME 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 procedure s.

8 -

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2284Q: 1 2-12

TABLE 2-1  !

STEAM GENERATOR DESIGN DATA (PER STEAM GENERATOR)

Original Replacement Design Pressure, Reactor Coolant / Steam i psig 2485/1085 N.C. j Reactor Coolant Hydrostatic Test Pressure (tube side), psig 3106 N.C.

l Hydrostatic Test Pressure, Shell Side, psig 1356 N.C.

Design Temperature, Reactor Coolant / Steam, OF 650/556 N.C.  !

Steam Conditions at 100% Load, Outlet Nozzle:  !

Steam Flow, lb per hr 3.31 x 10 6 N.C.

Steam Temperature, OF 521.2 N.C.

Steam Pressure, psia 821 N.C.  !

Feedwater Temperature at 100% Load, F 435.7 N.C.

Overall Height, ft-in 63-1.6 N.C.

l Shell OD, upper / lower, in. 166/127 N.C. l Shell Thickness, upper / lower, in. 3.5/2.6 N.C.

U-tube OD, in. 0.875 N.C. l Tube Wall Thickness, (nominal) in. 0.050 N.C.

Number of Manways/ID, in. 3/16 N.C.

Number of Handholes/ID, in. 2/6 6/6 ,

Number of U-tubes 3260 3214 Tube height (largest U-bend), in. 397.5 N.C. f Total Heat Transfer Surface Area, ft 2 43,467 44,430 I Reactor Coolant Water Volume, ft 3 945 925 ,

! Reactor Coolant Flow, gpm 89,000 N.C.  !

Secondary Side Volume, ft 3 4580 4682  :

Secondary Side Mass No Load, lbs 134,000 136,000 Secondary Side Mass 100% Power, lbs 89,000 91,000 [

Center of Gravity (from the support pads),

O' ft/in. 25-3.6 N.C.

O 2284Q:1

i TABLE 2-2 STEAM GENERATOR MATERIALS Original Replacement Plate (shell courses) SA-302 Grade B SA-533 Grade A Class 2 Tube Sheet Forging SA-336 Code Case 1332 SA-508 Class 2a Channel Head Casting SA-216 Grade WC SA-216 Grade WCC Support Plates SA-285 Grade C SA-240 Type 405 Channel Head Cladding Stainless Steel Stainless Steel, Type 304 or equivalent Type 304 or equivalent Tube Sheet Cladding Inconel Inconel Weld Deposit Tubes SB-163-61T (Code Case SB-163 Special Thermal 1336) Treated (Code Case 1484) i i

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l TRANS1 TION CONE o TRUhl0N -$ g O'  : e- LOWER SHELL BARREL QUATREFOIL TUBE f SUPPORT PLATE (6) - -

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CLOSURE F I NG - 450 400

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3.0 COMPONENT REPLACEMENT PROGRAM AND PROCEDURES p

d This section discusses the engineering evaluation of the field activi-ties required to implement the steam generator repair. Figure 3-1 represents the outage sequence, and Figure 3-2 illustrates the removal p sequence of the lower assemblies. These evaluations demonstrate feasi-I bility of implementation. Any changes incurred during detailed design will not alter the envelope of potential construction incidents dis-cussed in Section 5.2.

The steam generator lower assemblies will be removed and replaced through the equipment hatch in the containment structure. The upper assemblies will remain inside the containment and will likely be stored at the operating floor level, elevation +66', until the new lower assem-blies are brought in and installed.

Handling of the steam generator lower assemblies inside the containment will require an additional polar crane trolley with a capacity of approximately 250 tons. Handling through the equipment hatch will 8 require special track-mounted upending /downending skid attached to the steam generator vertical support lugs. (Ref. Figure 3-2) Handling outside the. containment is expected to be by heavy lift crane. Jacking operation may be used to minimize the use of heavy. lift crane.

A rigging platfom will be provided inside the containment. This plat-fom will provide support for the lower assemblies while they are being maneuvered from the vertical position to the horizontal position. The temporary construction loads from this platfom will be transmitted to

\s the containment base mat. In addition, temporary laydown area will be provided for two upper assemblies weighing 105 tons each and for two sets of swirl vane assemblies weighing approximately 5 tons each. Ade-p quate laydown area is available inside the containment, utilizing in h some cases temporary structural members to span certain operating floor l areas.

p.s Clearances to accomplish the removal of the lower assemblies through the l \v) l 2563Q: 1 3-1 1

equipment hatch may require the removal of minor quantifies of internal concrete and structures in the vicinity of the equipment hatch. The (m

portions to be removed are minimal and are listed in Section 3.2.5.

Impact on existing equipment is minimal and is described in Section 3.2.

m The removal of the lower assemblies through the existing equipment hatch will have minimal impact on the site layout in tems of new foundations requi red.

Upon exiting from the equipment hatch, the lower assemblies will likely be loaded onto a crawler type or rubber wheeled transporter and moved to an onsite temporary storage building.

The existing access to the equipment hatch will be modified as necessary to enlarge the access area and to facilitate access by trucks and mobile cranes. This access area will be at the equipment hatch elevation.

This refurbishment will provide additional flexibility in selection of handling methods for steam generator movement.

No pemanent modifications to existing structures are expected to be necessary.

3.1 PATHWAYS AND CONSTRUCTION RESTRICTIONS 3.1.1 SITE PREPARATION p 3.1.1.1 FOUNDATIONS Q ,1 All heavy hoisting equipment will be located so that foundations will not interfere with pemanent plant installations, either above or below i

(~] grade. The steam generator removal rails outside the equipment hatch V will require the placement of foundations as shown in Figure 3-3. Below grade emplacements may be removed after completion of the repair work.

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2563Q: 1 3-2

3.1. 2.1 TEMPORARY HEAVY LOAD BEARING PLATFORMS INSIDE CONTAINENT b(. A temporary platfom at elevation +24'-0" inside the equipment hatch opening will be provided to handle the steam generators as shown in Figure 3-3. The existing steel platfom in this area will be removed.

m Guide rails on this structure will extend through the hatch opening and align with similar guide rails outside the equipment hatch providing a path for the roller assemblies required to pemit steam generator move-ment through the hatch.

Additionally, there will be a platfom at elevation +66'-0" to support the polarcrane center post and for inverting the upper assembly. These are designed for adequate protection of the CRDM and the reactor under-neath.

3.1.2.2 OTHER PREPARATIONS Concrete removal and equipment relocation for rigging clearances are discussed in Section 3.2.

3.1.2.3 LAYDOWN FACILITIES INSIDE CONTAINENT l All fuel asemblies will be removed from the reactor and stored in the spent fuel pool. The reactor internals will be stored in the reactor vessel, the reactor vessel head will be stored in place on the reactor l Vessel and the control rod drive nechanism (CRDM) missile shield will be stored on top of the refueling cavity. Laydown of each upper assembly 1

's will be on a platfom at elevation +66'-0" in front of the steam genera-l (b tor cubicles. This platfom area will be evaluated and suitably strengthened if necessary to accomodate the steam generator upper assem-l bly loads.

O G' There is adequate space inside containment for the temporary storage of the additional equipment discussed in Section 3.2.

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l i 2563Q: 1 3-4 i

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3.1.1. 2 ROADWAYS, RAMPS, AND PLATFORMS The existing access to the equipment hatch will be modified as neces-sary. A service area shown in Figure 3-3 may be installed to facilitate access into the containment. The steam generator removal rails will be designed to support the weight as the lower assembly and associated V rollers and saddles. The service area will be equipped with as required loading and unloading of miscellaneous equipment.

p The proposed onsite steam generator haul ruute is shown in Figure 3-4.

The route shown will be used for both the original steam generator lower assemblies and the replacement assemblies. The haul route may vary based on detailed engineering studies. Prior to use of any haul route for transportation of heavy loads (other than nomal axle loads for highway equipment), the haul route will be evaluated for adequate capac-ity and upgraded where necessary utilizing standard construction prac-tices.

3.1.1.3 PROTECTION OF BURIED FACILITIES 8 Evaluations of the steam generator haul route and the potential for l

impact on safety related facilities are provided in Section 5.2.

l

! 3.1.1.4 STEAM GENERATOR RECEIPT ON SITE The method of transportation to the site will be overland by a multi-l axial rubber tire transporter from a intemodal transfer point.

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3.1.1. 5 LAYDOWN FACILITIES OUTSIDE CONTAINENT Adequate laydown area for construction materials and equipment will be provided by grading an area adjacent to the plant fence which was used V during original plant construction (see Figure 3-4).

3.1.2 CONTAINENT PREPARATION

' b) v i

2563Q: 1 3-3 l

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3.1.2.4 CONTAINENT STRUCTURAL ANALYSES

/

() Containment structural analyses have been perfomed in accordance with the design criteria in Appendix 58 of the FSAR for the following:

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

(See Figure 3-3).

p B. Containment base mat and existing floor support embeds in the con-tainment wall . A 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 pemanent modifications to the existing structures.

3.1.3 TRANSPORTATION ON-SITE ,

t 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.

l l q 3.1.4 RIGGING CONFIGURATION l

l 3.1.4.1 Inside Containment n The existing polar crane bridge will be analysed and strengthened to i

make it structurally adequate to sustain the loads imposed by a lower

! assembly in addition to a construction hoist weighing approximately 20 tons. The polar crane will be center posted to transfer part of the l

g lif t loads at EL. 66'-0. This will reduce loads on containment wall l Q brackets.

l 2563Q:1 3-5 l

l

Because of insufficient lift capacity, the existing polar crane trolley

(~3 is not suitable for steam generator lower assembly replacement, and U will, therefore, be moved aside to pemit the placement of a 250-ton construction hoist on the polar crane bridge. The temporary hoist will be load tested to meet current OSHA Safety Standards prior to its use p for construction lifts.

V The upper asserablies will be parted from the lower assemblies and lifted and inverted by pad eyes and commert:ial 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 compartraents using con-ventional hoisting techniques. The hoist lower load block will be linked by pins to a steam generator lift beam equipped with toggle ams or endless grommet type cables. The toggle ams 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, 8 such as Hillman roller units and structural members, will be used to move the assembly from the vertical to the horizontal position. Trans-f er 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.

3.1.4.2 OUTSIDE CONTAINENT Q The lower asserably will exit the containraent approximately at grade, on

\'v'/ 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 V Shipping saddles and tie downs will be provided for secure attachment while the transport device is in transit to and from the storage area.

l \

O 2563Q: 1 3-6

3.1.5 RIGGING AND HANDLING CONTROLS n All lift cranes and transport devices will be controlled such that

)

postulated failure does not adversely impact the ability to achieve and maintain safe-shutdown conditions in the operating unit or to provide adequate cooling for stored spent fuel. Administrative controls will limit lif t heights so that loads will be raised only to a height suffi-( cient to provide clearance for horizontal movement. When traversing plant roads in the vicinity of the operating unit, crane booms will be in the lowered position. Travel speed and travel routes for cranes and O transport devices will be controlled to minimize their influence on b) structures in the immediate area.

3.2 EQUIPENT AND CONCRETE REMOVAL AND REPLACEENT Engineering evaluations will be continued to detennine the impact of repair activities on equipment and structures in containment. The repair activity is not expected to result in any safety considerations due to equipment removal or interruption of function, nor will there be any major impact on structures and equipment (non-steam generator related) . ,

Detailed engineering studies are in progress to precisely define the components, pipes, cables, instruments, etc. within the containment affected by the repair activity. The discussion below provides the results of the study to date. It is provided to illustrate the minimal impact on non-steam generator related equipment within containment.

3.2.1 ECHANICAL EQUIPENT All equipment which interfer with the lower assembly pathway will be l 7_

temporarily removed and relocated as required.

/ \

As appropriate, equipment within the containment will be covered to ensure cleanliness during the repair.

l O

V 2563Q: 1 3-7 l

Upon completion of the repair, affected equipment will be returned to g service using standard procedures followed during routine plant mainte-V nance programs.

Disconnection of power cables to equipment is discussed in Section 3.2.3.

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3.2.2 INSTRUENTATION All steam generator instrumentation, sensing lines, and associated supports will be temporarily removed and relocated inside the contain-ment, as required.

The open ends of lines will be capped to ensure cleanliness during the repai r.

i The instrumentation and sensing lines will be returned to service using standard procedures followed during routine plant mainter.ance programs.

3.2.3 CABLE AND CONDUIT The steam generator repair does not require the removal and relocation of major pieces of electrical and control equipment such as panels, load f centers, transfomers or motor control centers.

The cable tenninations will be disconnected and the cables pulled back and coiled out of the path af the lower assembly. The same cable will then be reconnected when equipment are returned to their original loca-p tion.

The conduit to be removed will be tagged, disassembled, and the associa-ted cable pulled back and coiled. When the conduit is later rein-stalled, the cable will then be repulled and reconnected. Procedures I h will be generated for pulling back, coiling and repulling of cables and removal and reinstallation of the conduit. Circuit checkout procedures l will also be written.

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l 2563Q: 1 3-8 l

3.2.4 PIPING m

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.

b D. Steam generator blowdown piping.

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

Piping weld end preps, welding and nondestructive examination for the reinstallation will be in accordance with the latest applicable edition of the ASE Boiler and Pressure Vessel Code. The piping system will be 8 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".

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

T V C. A portion of the floor and removable floor slabs at elevation 21'-0" and 26'-0".

D. The upper portion of the steel stairway near the equipment hatch

'h opening.

2563Q: 1 3-9

E. A reinforced grouted pad in the equipment hatch at elevation +21'-0".

C

( F. A portion of the truss system tie rods to allow for clearance of the temporary polar crane trolly if necessary.

p 3.2.6 REMOVAL AND INSTALLATION OF STEAM GENERATORS V

3.2.6.1 GENERAL This section contains a general description of the removal and installa-hp tion in the lower steam generator assemblies and activities associated therewith. The discussion is limited to the methodology to be used and the basic removal and installation criteria. Detailed engineering is currently in progress to establish the specific techniques, processes, equipment, material, etc. which is required.

The basic technique which will be used to repair the steam generators will be to cut the currently installed steam generator at the steam drum upper girth weld. The inlet and outlet reactor coolant piping, the 8 steam line piping and feedwater piping will also be cut. The steam generator upper assembly will be lifj;ed off inverted and placed in a storage-work location in the containment for refurbishment of moisture i separation equipment. The lower assembly will then be lifted from its supports and transported out of the containment through the equipment hatch. The replacement assembly will proceed through the same steps in reverse. The upper and lower assemblies will be welded together in the i field. A detailed description of the process is given in following l A sections.

N) 3.2.6.2 GUIDELINES AND CRITERIA p A number of guidelines and criteria are applicable to the overall repair V process. They are sumarized below:

1. The reactor vessel will be completely defueled prior to the repair wo rk. The fuel will be stored in the spent fuel pool for the dura-x) 2563Q: 1 3-10 ,

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

p 2. Access to the containment will be through the present equipar 't and h personnel hatch; therefore, no structural changes will be required to the containment structure which foms part of the containment pressure boundary. Minor structural changes, e.g., chipping of concrete, may be required on internal walls; however, the effect on internal structures is expected to be insignificant.

3. The entire repair process will be preplanned. The guidelines contained in Regulatory Guide 8.8, "Infomation Relevant to Ensuring That Occupational Radiation Exposure at Nuclear Power Stations Will Be as Low is Reasonably Achievable", will be followed where applica-ble. In keeping with these guidelines mockups and training will be used to minimize outage time and radiation exposure. Decontamina-tic.7 and other exposure limiting techniques will be used where they 8 offer a significant savings in exposure commensurate with overall program objectives. Special scaffolding and other components will be prefabricated to the extent possible to minimize radiation expo-sure and outage time.

4. The reactor cavity will be covered by structural members to minimize the possibility of impacting the reactor vessel and associated components during the repair program.

[

5. The repair program will be completed in accordance with the Point I beach Quality Assurance Manual and Section XI of the ASE Code, I including such items as interaction of repair activities with the 1

j unaffected part of the plant station, design reviews, radiation ,

'\ control procedures, document control, material acquisitions, etc.

6. The actual repair process will be similar to the methods used during original construction of the units. Much of the experience gained

\j 2563Q: 1 3-11

during original construction is applicable to the repair process and will be used as appropriate.

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7. The potential environmental effects of the repair program are expected to be minimal. However, reasonable precautions will be exen:ised to further minimize any environmental impact.

i p

8. 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 Q of special concern are facilities to prevent the spread of radio-active contamination, disposal of radioactive material, and security provi sions.

9. The major portion of the repair program will be perfomed by a comercial installer under the direction of Westinghouse personnel.

It is presently anticipated that WE will utilize its own radiation control procedures and personnel. The installer wili provide quality control personnel and procedures and Westinghouse will 8 provide quality assurance personnel. The installer will be required to have an ASE certification as, applicable to the work he is to l perfom.

10. The length of the steam generator repair outage is now estimated to be approximately 180 days. This schedule is predicated on perfom-ing in containment work on shift rotation that will pemit working 24 hrs / day. The schedule is divided into the following phases:

O V a. Preshutdown activities

b. Shutdown and preparatory activities

c. Removal activities q d. Installation activities b e. Post Installation activities

f. Startup activities

g. Post Startup activities O

2563Q: 1 3-12

Each of these phases is discussed in the following paragraphs.

l}

V 3.2.6.3 PRESHUTDOWN ACTIVITIES Prior to the first unit shutdown, the repair program will be preplanned m and appropriate provisions made for accomplishing each activity required 1

v' 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 Q items which do not require a unit shutdown.

The " work package" concept will be used for the repair program whereby individual taske will be defined and a work package for each task, containing all pertinent infomation required to complete that task, will be completed.

3.2.6.4 Post Shutdown Activities 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 perfomed; however, they are not necessarily listed in the order which they will be perfomed.

1. Establish appropriate valve linups consistent with the requirements of the repair efforts.

(3 V 2. Place systems in the appropriate condition for long tem layup, i .e. , approximately six (6) months.

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

(d Remove reactor pressure vessel head and upper internals and store.

4.

5. Remove all fuel assemblies from the reactor vessel and store in 2563Q:1 3-13

spent fuel storage pool in the fuel building.

b' d 6. Reinstall reactor vessel internals, and replace reactor vessel head.

7. Survey containment work areas and establish radiation zones.

O V 8. Perfona local decontamination of work areas to the extent possible, and shield areas which cannot be decontaminated.

9. Install cover over the reactor cavity to provide protection to the j reactor vessel and associated equipment and to provide a continuous work area. The cover or flooring will be supported by structural members designed to hold the center post of the upgrated polar c rane.

10. Assemble special prefabricated scaffolding to gain access to all work areas.

11. Remove biological shield wall and transport debris from the con-8 tainment. These walls are not structural members, they are only required for shielding. ,

12. Remove insulation from steam generators, feedwater piping, steam line piping, reactor coolant piping, and other components and l

transport debris from the containment.

13. Install local control structures, such as tents ducting, temporary filters, etc.

v

14. Install the steam generator transport system, e.g., rails, inside the containment and through equipment hatch.

O Refurbish equipment hatch acress area and install steam generator C/ 15.

removal rails outside containment.

16. Inspect, test, and modify the existing containment polar crane as

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2563Q: 1 3-14 l

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

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

Oi 1 V 3.2.6.5 REMOVAL ACTIVITIES Having established appropriate access requirements, radiological n controls, installation of temporary facilities to provide access to the b work areas and for removal of debris and components, and the removal of insulation from the equipment, the actual removal process chn comence.

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

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 generators will be perfonned roughly simulta-8 neously; however, because of availability of the polar crane, the cocraencement 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.

i I

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

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

l 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

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V the upper and lower shells can be lifted. The removed section will be marked for identification and stored for reuse. (See Figure 3-5)

3. Cut feedwater piping at its junction with the upper shell and d

2563Q: 1 3-15

upstream from the junction to allow a section of the piping to be o removed so that the upper and lower shell can be removed.

(See Figure 3-6)

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 sections, 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 which will be removed.

5. Cut steam generator wrapper to facilitate lifitng 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 8 machi ni ng. The lower assembly will not be reused; however, the shell of the upper assembly wil) be used.

7. After removal of the upper assembly, it will be placed in conveni-ent location within the containment in the inverted position, i.e.,

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

p) 8. The steam generator lower assembly will be lifted from its V 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 lifting trunnions on either side of the steam generator. A ,

special lifting rig must be used to attach these cables to the d crane hook. As shown in Figure 3-2, the steam generator will be lifted straight up out of its supports, moved aside in the vertical position to a designated location on the operating floor, and lowered onto upending /downending skid in the vertical O

2563Q: 1 3-16

position. The lower assembly will be downending and vertically transported out of the containment through the equipment hatch on d rails.

9. Radiological controls will be in effect during the removal process. Since the lower assemblies will be a major soun:e of radioactive contamination cutting operations will be controlled and covers will be employed to control the spread of contamination.

10. During all activities, the containment work areas will be cleaned

) and decontaminated as required.

v

11. Following the transport of the steam generator lower assemblies through the equipment hatch, they will be transported to temporary storage facility onsite.

3.2.6.6 Installation Activities Following the removal of both steam generator lower assemblies, the 8 major installation activities will commence. The major steps involved in the installation process are disc,ussed below.

1. The replacement steam generator lower assemblies will be delivered to the site by rubber tired transporter. While it is planned to install the steam generator lower assemblies at or near their time of delivery, it may be necessary to store the steam generators onsite for a short period of time. Provisions will be made for

,-s appropriate storage.

V) 6

2. The steam generator lower assemblies will be delivered to the equipment hatch and lifted on the rails for transfer into contain-ment. The procedure will be the reverse of that for removal of Qi the lower assemblies.

3. The assembly will be transported to a designated location within the containment using the special lifting rig with straps or O

V 2563Q: 1 3-17

cables attached to the two lifting trunnions on the replacement lower assembly and the assembly will be upended using the polar d c rane. The assembly will then be lifted vertically and moved to a position over the steam generator supports. The assembly will be lowered onto place in the supports. Temporary positioning devices, (e.g. Jacks) may be installed to facilitate the position-V ing of the lower assembly.

4. Reassemble and/or reinstall the steam generator support system.

5. Install new moisture separation equipment, feedring and other internal components in the upper shell. Weld preparation for the upper shell will be made for mating to the lower assembly.

6. Lift upper assembly into place and align with lower assembly.

Temporary positioning devices may be used to facilitate alignment.

7. Weld the upper and lower assemblies together,' stress relieve and l inspect.

8. Weld the steam generator wrapper to the upper internals and I

inspect.

9. Reinstall the reactor coolant piping. The installation procedure for this pl ping is similar to that used during the original installation of the steam generator.

10. Complete fitup, weld, and inspect the main piping.

11. Complete fitup, weld, and inspect the feedwater piping.

12. Reinstall miscellaneous piping (e.g. blowdown) and other equipment d which was removed to provide clearance for lower assembly removal.

13. Reinstall instrumentation and controls which were removed.

O 2563Q: 1 3-18

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

15. Remove all temporary structures which were put inplace to facili-tate the repair process.

O Q 16. Restore crane wall and other concrete structures which were chipped.

3.2.6.7 POST INSTALLATION ACTIVITIES O Following the completion of the major installation activities, it will be necessary to restore the unit to a condition from which unit startup can comence, and to perfom tests or inspections. The following are typical activities which will be perfomed:

1. Perfom hydrostatic tests in accordance with Section XI of the ASE Code.

2. Clean affected systems and work areas.

3. Install insulation.

4. Remove scaffolding.

5. Remove cavity cover.

s 6. Refuel the reactor.

N

7. Perfom baseline inservice inspection as required on piping, equipment or components, including 100 pen:ent eddy current inspections of steam generator tubing.

8. Remove all temporary structures and supports.

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

2563Q: 1 3-19

3.2.6.8 STARTUP ACTIVITIES Upon completion of the repair work, a number of activities must be completed to return the unit to power. Among these are:

1. Establish valve lineups and system conditions in accordance with l O

V established procedures.

2. Perfom startup test program on the systems that were affected by the repair effort or otherwise required to satisfy test' program i requirements and the Technical Specifications.

3. Run perfomance tests and moisture carryover tests to verify perfomance of the steam generators.

3.3 RADIOLOGICAL PROTECTION PROGRAM The radiological protection program to be implemented for the repair effort will be in accordance with the Point Beach Nuclear Plant Health 8 Physics Administrative Control Policies and Procedures Manual. Thi s program is responsive to the applicable Nuclear Regulatory Commission l and State of Wisconsin regulations. l 3.3.1 SUPPLEENTAL ACCESS CONTROL i

Additional facilities will be provided for the repair effort to accom-modate the personnel involved. These facilities include:

l V A. Outside controlled zone l

1. Office area O

l (V 2. Storage area

3. Radiological protection training area l

V 2563Q: 1 3-20

4. Mock-up training area B. Inside controlled zone

1. Office area m

2. Radiation control point

3. Protective clothing alckup area s

4. Locker room

5. Protective clothing dressout area

6. Storage area for protective clothing

7. Sanitary Facilities ,

8. Health physics area The following is a brief description,of the access control pathway currently contemplated for entering and exiting the containment:

After proceeding through the radiation control point, personnel will pick up their protective clothing and dress in a locker room area before entering the equip-ment hatch into containment. The requirements for protective clothing are specified in the Point Beach i d Nuclear Plant Health Physics Administrative Control Policies and Procedures Manual HP 2.5, General Use of Protective Clothing and HP 12.2, Nonroutine Protective Clothing Requirements. Personnel leaving the contain-l V ment will remove their shoe covers and gloves at an access control point just inside the equipment hatch r before stepping onto the step-off pad. These person-nel will imediately proceed to the undressing area to a

2563Q: 1 3-21

remove protective clothing and be checked for residual contamination. They will then exit through the radia-J tion control point and return to the locker area for their street clothes. Handling of contaminated personnel will follow the procedures given in the l Point Beach Nuclear Plant Health Physics Administra-tive Control Policies and Procedures Manual HP 9.1, Personnel Decontamination.

Access control at the steam generator compartments, equipment hatch, and O

\g at the temporary access control area discussed above will be provided.

Personnel involved in work areas with a potential for high-level contamination will wear two sets of protective clothing. . The outer set of protective clothing will be removed when leaving the work area and deposited in a container. The second set will be removed in an area outside containment, as described above.

3.3.2 LAUNDRY In order to accomodate the laundry expected during the repair effort, additional water wash or dry cleaning capabilities will be provided.

Laundering of protective clothing and cleaning and sanitizing of respiratory equipment wil be in accordance with the Point Beach Nuclear Plant Health Physics Administrative Control Policies and Procedures Manual.

3.3.3 CONTROL OF AIRBORNE RADI0 ACTIVITY AND SURFACE CONTAMINATION Q

Airborne radioactivity inside containment during the steam generator I

repair effort will be controlled and monitored. Small releases may p occur via the plant ventilation systems. A slightly negative pressure Q inside of containment will be maintained using the containment purge exhaust system. Air will be drawn through the equipment and personnel hatches and exhausted by the purge system via the purge vent, thus precluding airborne radioactive particles or gases from leaving contain-()

2563Q: 1 3-22

ment openings utilized for construction activities. Air flow require-ments necessary for maintaining a slightly negative containment pressure V are well within the existing purge exhaust system capacity. The air being exhausted will be monitored for radioactivity using nomal plant monitoring system.

p C' In addition to bulk containment atmosphere control of airborne radio-activity, appropriate localized control will also be provided. Radio-activity generated during the cutting of the reactor coolant pipes will 7

be contained within specially designed contamination control envelopes,

( which will provide local high efficiency filtration. Personnel working inside these control envelopes will wear respiratory protection equip-ment, as required, by Point Beach Nuclear Plant Health Physics Adminis-trative Control Policies and Procedures Manual HP 2.6, Respiratory Protection and HP 12.1 Respiratory Protection Program, Appendices A through J. No special provisions are expected to be required for other cutting operations inside containment.

Section 3.3.1 describes the method of controlling the spread of surface 8 contamination by personnel removing their outer set of protective clothing when leaving the control envelope.

The radioactive releases and dose assessments associated with Steam Generator Repair are provided in Section 5.2.2.1.

3.3.4 SUPPLEENTAL PERSONNEL MONITORING REQUIREENTS

{

j C

N 3.3.4.1 MONITORING OF AIRBORNE RADI0 ACTIVITY l

Mobile air monitors will be used, as required, to monitor the airborne l

radioactivity inside the control envelopes and in other work areas inside containment. Airborne radioactivity samplers coupled with l

h laboratory analyses will also be employed.

3.3.4.2 MONITORING 0F WORKERS FOR INGESTED RADI0 ACTIVITY 2563Q: 1 3-23

3.3.4.2.1 Whole body counting of personnel shall routinely be done for those individuals exposed to significant amounts of radio-active materials. Whole body counting shall also be done after temination of work at Point Beach Nuclear Plant for all employees having access to the controlled side and after any significant event that may have contributed to an V uptake.

Whole body counting of contract workers shall nomally be accomplished during " processing in" if the individual has worked as a radiation worker elsewhere during the previous six months, and when " processing out" at the end of the work assignment if the work involves respirator use for radiation protection.

3.3.4.2.2 Nasal swabs will be taken from individuals when there is a possibility of accidental uptake. The results of analysis of these swabs will be reviewed for possible imediate whole body counting.

3.3.4.2.3 Urinalysis is perfomed to detect tritium and to detemine the existance of other beta gama emitters when compared to the whole body count.

3.3.4.3 PERSONNEL MONITORING All personnel working or visiting within the plant restricted area shall ,

be provided with either a themoluminescent dosimeter (TLD Badge), a self reading dosimeter (SRD), or a combination of both in accordance x with the Point Beach Nuclear Plant Health Physics Administrative Control Policies and Procedures Manual HP 3.1, Issue and Control of Personnel Monitoring Devices.

D (V 3.3.4.4 RADIATION AND CONTAMINATION SURVEYS Detailed surveys which provide proper control of radiation and contami-nation will be perfomed, as required, throughout the repair effort.

2563Q: 1 3-24

These surveys will be perfomed in accordance with Section 8.0 of the p Point Beach Nuclear Plant Health Physics Administrative Control Policies C and Procedures Manual HP 8.1, Contamination Surveys; HP 8.2, Radiation Surveys; HP 8.3, Postings of Radiation and High Radiation Areas; HP 8.4, Extended Outage Survey Schedule; HP 8.5, Airborne Radioactivity Surveys; and HP.6, Counting of Air Samples for Low Level, Long Lived Radioactive Q Particulate Contamination.

3.3.4.5 PORTABLE SURVEY INSTRUMNETS Table 3-1 provides a typical listing of the types of portable survey instruments which are used during the repair effort.

3.3.5 GENERAL ALARA CONSIDERATIONS The repair of steam generators in operating nuclear power plants requires the utilization of state-of-the-art exposure reduction tech-niques to keep radiation exposures As-Low-As-Reasonably-Achievable (ALARA). The experience gained by the nuclear industry from the replacement of six (6) steam generators at the Surry Unit 1 and 2 sites 8 and the three (3) steam generators replaced at the Turkey Point Unit-3 site will be used to the extent possible in the engineering, planning, tool and process designs, and exposure control for the Point Beach Nuclear Plant Unit 1 Steam Generator Replacement Program.

l Personnel exposures will be maintained ALARA in accordance with 10 CFR 20.1 (c) and the guidance provided by Regulatorj Guide 8.8. An exten-f l sive program to address radiological concerns has been established which l p consists of utilizing remote and semi-remote tooling, remote monitoring l b systems, extensive training of personnel in full size mock-ups, the utilization of temporar/ stdelding to minimuze the radiation field and the use of administrative controls to limit the number of non-essential personnel at the work station areas.

(/

3.3.5.1 SPACE ENVELOPE EVALUATION The ingress / egress accessibility and the effective control and utiliza-V 2563Q: 1 3-25

tion of in containment work spaces are essential for minimizing exposure p while working in potentially high radiation areas. The computer aided h design (CAD) computer system, modeling and scale drawings will be used to confim access clearances for the movement of tools and equipment in and out of containment. These techniques will minimize the potential for unexpected delays in containment work and the associated radiation

() exposure. Work space envelopes will be studied to assure adequate space for the tooling designed to be used in the high radiation environment.

3.3.5.2 TEMPORARY FHIELDING

/ s b Shielding will be used, as necessary, to reduce the dose rates from other components such as the reger:erative' heat exchanger, RHR . system valves, and from temporary storage areas used for storage of contamina-ted pieces of pipe, rags, and tools. Temporary shielding will be used, as necessary, for the steam generat'or while it is being cut out of the reactor coolant loop and while the steem generator and reactor coolant pipe is noved out of the containment. The steam generator shell will also help shield the more contaminated parts of the stean generators.

The water level on the' secondary sidp of the steam generator will also be adjusted as required to provide shielding during cutting of the upper shell and feedwater and steam lines.

3.3.5.3 LOCAL DECONTAMINATION Local decontamination of the steam generators and reactor coolant piping N may be perfomed. Decontamination of the work areas will be perfomed (h

periodically depending on the contamination levels. Paper and plastic sheeting will be used to facilitate collection and cleanup of contamina-tion.

N

{Q 3.3.5.4 LOW 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 (m\

G' 2563Q: 1 3-26

these areas. Signs will also be posted in high cackground radiation areas to warn personnel.

Health physics personnel will work with the job supervisors to assure that personnel not required in the work area remain in the waiting area.

O Q 3.3.5.4 RADIOLOGICAL PROTECTION PERSONNEL TRAINING As a minimum, personnel will be given radiological protection training as described in the Point Beach Nuclear Plar.t 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 additical 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 alams and response, and ALARA.

3.3.5.5 M0CK-UP TRAINING ,

The extensive training of repair personnel on full-sized mock-ups has proven to be an effective method for minimizing personnel exposures durng steam generator maintenance. Steam Generator Repair personnel will become thoroughly familiar with each tool that is designed and 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 V designs will be modified and procedures will be updated for field imple-mentation based on lessons learned in mock-up testing. l

~ In parallel to the field qualifications of the tooling, personnel will s

be trained and qualified to operate the equipment in field operations.

Technicians will be required to operata the equipment in simulated radiation conditions, dressed in the complete Anti-C clothing required n

V .

i 2563Q: 1 3-27

for perfoming the repair in operating plant conditions. An estimate of p time spent in the radiation environment will be used to detemine man-b power requirements and to establish administrative man-rem exposure limits for completing the repair. The training records of each techni-cian will be documented and fonvarded to the field coordinator so that only qualified personnel are used in the steam generator replacement V' 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.

3.3.5.6 REMOTE MONITORING SYSTEMS The utilization of remote monitoring systems will be used during in the repair of steam generators. TV cameras will be used to monitor work both inside and outside of the steam generator cubicle. 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 8 in the high radiation field.

3.3.6 MISCELLANE0US WASTE DISPCSAL 3.3.6.1 CONCRETE DISPOSAL Approximately 15 cubic yards of concrete will be removed from the containment internal walls and floors and will be disposed of. The 3 majority of this concrete has an insignificant amount of transferable (O contamination (transferable contamination is considered insignificant if it is less than 2200 dpm/100 cm 2 per 49 CFR 173.397) without surface decontami nation. The concrete which is considered contaminated, (i) may q 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 C burial site.

2563Q: 1 3-28

3.3.6.2 MISCELLANEOUS DRY WASTE DISPOSAL O

V Metal shavings from the various cutting operations and miscellaneous dry waste, such as paper, rags, etc., will be put in standard shipping containers and shipped as LSA material to a licensed land burial site, q The estimated volume of low-level waste is 760 m3 (26,800 Ft3 ) per D steam generator as indicated in USNRC NUREG-CR-1595. The total activity in this waste is estimated to be 21 curies, derived from the Surry Unit 2 data also reported in NUREG-CR-1595.

3.3.6.3 LIQUID RADWASTE DISPOSAL There are three potential sources of radioactive liquid to be disposed of. These sources are:

a. Water drained from the reactor coolant system

b. Laundry waste water

c. Local decontamination waste fluids.

The radioactive releases associated with these soun:es are discussed in Subsection 5.2.2.4.

The reactor coolant will be prccessed through the boron recycle system and will be released or reused, aepending on plant water inventory.

p The laundry waste water could be discharged without processing due to the low activity level as indicated by the estimated laundry waste water specific activities given in Table 5-8. However, the laundry waste water will be processed in the the nomal liquid radwaste processing system.

v The small amount of liquid waste gene-ated as a result of local decon-tamination will be processed in the nomal liquid radwaste processing sy stem.

d t

2563Q: 1 3-29

3.4 DISPOSITION OF STEAM GENERATOR LOWER ASSEMBLIES A

Cl The lower assemblies to be removed from Point Beach Unit 1 represent the single largest soun:e of solid radioactive waste to be disposed of during the repair effort. The disposal effort is independent of the repair and is evaluated on tnat basis.

g)

(

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

3.4.1 OBJECTIVES OF HANDLING / DISPOSAL OPERATIONS The objectives of handling / disposal operations are as follows:

A. Ta dispose of or store the lower assemblies safely and economically, 8 and in accordance with applicable licensing requirements.

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

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

L!

3.4.2 ONSITE STORAGE  :

A temporary onsite storage building will be provided for the storage of t ,/ the lower assemblies. It is expected that the lower assemblies will be stored in this building until plant decouaissioning. Prior to removal from the containment, the openings in the lower assemblies will be sealed by welding to prevent the release of radioactivity during trans-L/

2563Q: 1 3-30

fer and subsequent onsite storage. As discussed in Section 3.4.4, the Q only radiological consideration associated with storage is the direct U 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 p with onsite storage.

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 evaluations 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 i Table 3-2. With the exception of the immediate intact shipment alterna-tive, the long-tern onsite storage alternatives provide the lowest esti-v mated occupational exposures.

1 3.4.4 RADIOACTIVE RELEASES AND DOSE ASSESSENT ASSOCIATED WITH l p ONSITE STORAGE O i As indicated in Section 3.4.2, prior to removal from the containment, the openings in the steam generator lower assemblies will be sealed to g prevent the release of radioactivity during transfer and subsequent

\ )

l 2563Q: 1 3-31

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

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

l As discussed in Section 3.4.5, the radioactivity within the steam generators is imobile and the lower assemblies are stored in a closed l facility. Thus, even if seal integrity were lost, releases to the envi-ronment are not likely. Nonetheless, a surveillance program will be l

implemented. Periodic area radiation surveys and monitoring will

( provide assurance that there are no releases of radioactivity to the envi ronment.

The only contribution, therefore, to,the annual dose equivalent to any 1

I 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 individual would be continuously exposed for a period of one year at the V site boundary; therefore, the actual annual dose equivalent to any indi-vidual at this location will be substantially lower than that given above.

m 3.4.5 ACCIDENT CONSIDERATIONS ASSOCIATED WITH ONSITE STORAGE The only potential accident consideration associated with steam genera-tor lower assembly storage is the release of radioactivity to the i k.) .

2563Q: 1 3-32

l envi ronnent. The majority of this radioactivity is on the primary side surfaces of the lower assembly in the fom of a protective film of raetal 73

() oxides which is very adherent and very refractory. Radioactivity would be present in negligible concentrations on the secondary side of the steam generator.

. ,a As discussed in Section 3.4.2, an additional measure of radioactivity l

{)

confinement will be attained by welding cover plates over all lower asserably openings.

(O i A. Radioactivity could conceivably be released to the environuent only

'J l

if both of the conditions below occurred:

1. Radioacti fity is dislodged froa the primary side surfaces.

2. The lower assembly priraary side boundary is breached.

B. There are three mechanisms which could potentially dislodge the corrosion film:

1. Thernal shock. ,

2. Chemical / corrosion attack.

3. Mechanical shock.

Temporary variations in the lower asseubly would occur during ambient p temperature variations, but these are much too slow to produce a themal

() shock effect. Since the lower assemblies will be drained and sealed against moisture, chemical and corrosive attack will not occur. The possibility of mechanical shock during storage is very small since the steam generators are protected by the closed temporary storage facil-g

() i ty. Even if a aechanical shock is assumed, the tenacious nature of the filra is such that it would not dislodge more than an insignificant amount of radioactivity and, even then, any dislodged radioactivity would be contained in the sealed steara generators.

(q) v 2563Q: 1 3-33

Since it is highly unlikely that more than an insignificant amount of radioactivity would be dislodged from a primary side surface, the second d condicion for radioactivity release to the environment, breaching the i

lower assembly priaary side boundary, need not be considered.

p Based on the above, it is concluded that there are no radiological acci- ,

(,) dent consicerations associated with onsite storage.

3.

4.6 CONCLUSION

S O The steau generator lower assemblies will ultiaately be disposed of at a Q

licensed land burial site or decocaissioned with the plant. Radiologi-cal iapacts associated with this disposal alternative are acceptable and are less than those associated with presently available alternatives.

3.

4.7 REFERENCES

FOR SECTION 3.4

1. Hoenes, G. R. , Mueller, M. A. , McComack, W. D. ,1980. Radiological Assessment of Stean Generator Removal and Replacement: Update and 8 Revi sion. NUREG-CR-1595 (PNL-3454), U. S. NRC, Washington, D.C.

)

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2563Q: 1 3-34

3.5 PLANT SECURITY (O

( ,) Specific plans for physical protection of Point Beach Nuclear P1 ant 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).

(mj' 3.6 QUALITY ASSURANCE PROGRAM The Quality Assurance Programs for WE and Westinghouse Electric Corpora-( ) tion are described in this section.

3.6.1 WE QUALITY ASSURANCE PROGRAM Wisconsin Electric Power Company (WE) has the overall responsibility for the Quality Assurance Program for replacement of steam generators hi accordance with Appendix H of the Final Safety Analysis Report.

WE will assure that Westinghouse has documented and implemented Quality 8 Assurance Programs comuensurate with their scope of work and in accord-ance with the requirements of 10CFR50 Appendix B and the ASE Code.

Quality Assurance (QA) and Quality Control (QC) functions will be perfomed on site by Westinghouse and its subcontractors under QA programs approved by WE. WE QA personnel will audit and provide surveillance to the extent necessary to assure that all activities are l

conducted in accordance with applicable codes, standards and regulations and are in concert with the WE QA program.

1 p)

\_

l 3.6.2 WESTINGHOUSE NUCLEAR SERVICE DIVISION (WNSD) QUALITY ASSURANCE l

PROGRAM n

, ) The quality assurance program used by WNSD during the installation of the replacement steam generators will be in accordance with " Westing-l house Nuclear Service Division Quality Assurance Program Plan,"

WCAP9245, Rev. 5, May, 1980.

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L]

l l

2563Q: 1 3-35

3.6.3 WESTINGHOUSE NUCLEAR TECHNOLOGY DIVISION (WNTD) QUALITY ASSURANCE PROGRAM O

The design of the replacement steam generators is in accordance with the quality assurance progran described in WCAP-8370, Rev. 9A, Amendment 1,

( " Westinghouse Nuclear Technology Division Quality Assurance Program."

3.6.4 WESTINGHOUSE NUCLEAR COMPONENTS DIVISION QUALITY ASSURANCE PROGRAM The quality assurance program for the fabrication of the replacement i steam generators is in accordance with " Westinghouse Nuclear Coaponents Division Quality Assurance Program Manual," Rev. 5, dated 4/27/82.

t l

l

< i 1

O  :

i i

2563Q: 1 3-36

O O O O F O O TABLE 3-1 1 of 5 TYPICAL PORTABLE SURVEY INSTRUMENT iPECIFICATIONS Power Type of Application at Point Beach Instrument Detector Source Radiation Range Controls Nuclear Plant Measured Eberline Model Air-filled toniza- Two "D" size Beta- 5,50,500 mR/ int; One zero check pushbutton. General purpose survey R0-1 tion chader, with cell s , 1.0-1. 7 gamma 5,50,500 mR/hr; One zero set knob. One instrument beta shield for window volts each 5,50,500 R/hr; range switch with 0FF and full scale + 101 battery check positions.

Eberline Model Air-filled toniza- Three NRDA Beta- 5,50,500,5000 One rotary switch with General purpose survey R0-2 tion chaser, with 1604 (9V) gamma mR/hr DiF, battery check, zero instrument beta shield for batteries -+ 5% check and range selection window positions. One zero set knob.

Eberifne Model Air-filled f oniza- Four miniature Beta- 50,500 mR/hr One rotary switch for the General purpose survey R0-3A tion chamber, with NEDA 1604 (9V) gamma 5,50 R/hr OFF, battery check, zero instrument beta shield for batteries -+ 5%

set, and scale selector.

window One zero set knob.

Victoreen Ionization chaser Six 22-1/2 volt Gamma 10,100 & 1000 One zero adjust switch. Emergency radiation Model 592-B with no beta window batteries and mR/hr + 10%

One selector switch with monitor three 1.3-volt fivc positions for zero mertury cells set and scale multiplica-tion.

Eberline Model Gas-filled ionization Two NEDA 1604 Beta- 1-1000 mR/hr One control switch turns General purpose gamma PIC-6A chaser with beta win- (9V) batteries gamma 1-1000 R/hr instrument OFF, provides radiation survey dow on bottom (60 cm. -+ 20 %

battery check and selects instrument Hg pressure-propane) the range

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

)

I j TABLE 3-1 (Continued) 2 of 5 i

TYPICAL PORTABLE SURVEY INSTRUMENT ' SPECIFICATIONS j Power Type of App 1(cation at Point Beach '

j Instrument Detector Source Radiation Range Controls Nuclear Plant ,

, Measured  !

} [berif ne Model Ni ne-i nc h-diameter Five standard Neutron 5,50,500 One selector switch for General purpose neutron +

1 PNR-4 cadnlum loaded, poly- "D" size cells 5000 mR/hr 0FF, ON, and BATT. One radiation survey instrument

! lene sphere with a full scale + 101 H. V. adjust.  !

etptube BF in the center -

l Eberline Model Two C-M tubes inside a Four "C" size Beta- 2,50 ar/hr A One selector switch for High level radiation monitor i 6112 telescoping probe that cells gamma 2,50,1000 R/hr 0FF ( AUS), battery and used in areas which are not

extends over thirteen ~+ 105 scale changes. One easily accessible >

} feet- connection for Aural j indication.

Eberitne Model Nine-inch diameter One 6.3-volt Neutron 10 mR/hr to One reset for Hi-Lo Special neutron surveys Re-16 cadnf um loaded, poly- gelled electro- 10 R/hr alaru. One switch for ,

power DN. One pushbutton lite battery etityjene sphere with -+ 10%

a BF tube in the with trickle for BATT. check. Has center charge connec- connection for scaler. '

tion for AC Victoreen Neher white ionization Four standard Gasmia 0.1-100 mR/hr One rotary switch for 0FF, High radiation surveys and j Radector Ill chast>er "D" size cells, 0.1-100 R/hr battery and logarithmic emergencies

NEDA type 13 0.1-100 kR/hr scale. One pustd>utton i

~+ 20% switch for scale illumi-l nation.

1 i

t 4

)

i 2563Q:1 _ _ _ _ ,,__ , , , _ _, , _ _ _ , _ . _ . _ . _ . _ _ _ _ _ . _ _ , , _

o 4

l C o o e

i' TAKE 3-1 (Continued) 3 of 5 I TYPE AL PORTABLE SURVEY INSTRlNEN1 SPEC IFE ATIONS j Power Type of Application at Point Beach Instrissent Detector Sourte Radiation Range Controls Nuclear Plant Measured Eberline14odel G-M detector " side One Ni{ d bat- Beta- 0-500, 0-5000 One selector switch for Used in monitoring of Re-14 window" with beta tery with tric- gamma 0-05,000 cpm 0FF, BATT. and scale mult- equipment and personnel shield kle charge con- ~+ 5% p11 cation. One volume nection f or control for Aural indica-105-125V E tion. One reset switch.

One alarm set switch (in back).

G-M detector "end Same Be ta- Same Same Same window" with no beta gamma shield HP-210 G-M detector Same Beta- Same Same Same "Pa ncake" ganna Eberline Portable G-M gamma detector Five standard Ganna 0-2 R/hr Scale multiplier switch High energy ganna monitoring Alpha C ounter "D" size cells ~~+ 10%

and 0FF, external detector Model PE-15AGA switch & two discriminators E-3 alpha scintil- Same Alpha 0-2000 cpm Same Alpha monitoring lation detector + 10%

PG-1 plutontisa gamma Same Gas:saa 0-2 R/hr Same For information only scintillation detector + 10% (not used) 2563Q:1 1

f

. , . . . - , - , - . - _ . . . _ _ _ , - , _ , , . - . - - . ~ , - - . . . .- .,...---. ._ _ _. ...m -. . _ , , , , , - . . . . . ..,-.,._s , . . - - , . _ . - ,- - , , - . - , ,

_ _ _ _ . - , _ _ . _ . - - - - - --- _- - - - _. -m.-- - - - - - -- - -- -- --

4

$ \

I i

i i

TABLE 3-1 (Continued)

TYPE AL PORTABLE SURVEY IN5iRlHENT SPECIFE ATIONS 4

} Power Type of Appitcation at Point Beach j Instrument De tector Sourte Radiation Range C ontrols Nuclear Plant 1 Measured 1

q Eberline Model HP-210 G-M detector One Ni-Cd Beta- 0-500, 0-5000, One selector switch for Counting high level smears i RM-19 "Pa nc ak e" battery with gasuna 0-50,000, 0- 0FF, BATT, and scale mul- and as frisking unit in

trickle charge 50,000 cpe, plication. One volume field for checking person-connection for -+ 155 control for Aural indica- nel and equipment

' 105-125V AC tion. One reset switch.

One high voltage push to read switch. Gross /PHA selector switch.

I I Eberline Model RP-1/78 C -d detec tor Five standard Beta- 1-500 K cpm One selector switch for Used for detecting con-

] 1 IM " side windcw" with "D" size cells gamma '~+ 105 0FF, ON, and 8ATT. One tamination, beta-gasosa.

beta shield H. V. adjust.

• Johnson Model Johnson GP-200 "end ree standard 0-20 mR/hr Range switch off X1, X10, Used solely as a count rate i GSM-5 window" "D" size ergis 0-500 cpm and X100 hattery test and instrument

, + 155 calibration s

HP-210 G-M detector Same Be ta- Same Same Same "Pancak e" .i gasuna l

1 Victoreen Thyac Model 489-4 beta- Two standard Beta- 0-0.2 mR/hr Rotary range and function Use limited to hospital III, Model 490 gasssa probe "D" size cells, gassaa 0-2 mR/hr and 0FF, battery. X1000, emergency room, site boundary NEDA Type 13 0-20 mR/hr X100, X10 and XI cpm and control center. First aid i

+ 10% 0-2, 2, and 20 mR/hr. station and health physics Rotary response switch station marked slow, medium, and l fast.

t i

e  !

2563Q:1 i l

]

i c , , ,

o 1

TABLE 3-1 (Continued) 5 of 5 TYPE AL PORTABLE SURVEY INSTRtHENT SPEC IFE ATIONS j Power Type of AppItcation at Point Beach l

Instrument Detector Source Radiation Range C ontrols Nuclear Plant i

Measured I

i t

Victoreen Thyac Model 489-35 alpha- Same Alpha- Same Same Same I 111 Model 490 beta-gamma probe beta-ganaa l Eberitne Fast. Model NP-2, BF3 slow Five standard Slow & 500;.5000;.50,000 Switch for OFF, ON, and Special neutron surveys

] Slow Neutron neutron tube with base "D" size cells fast and 500,000 c'pe battery check. HV adjust.

i C ounter Model assembly neutrons ~+ 10%

j PE -4 l

Dosimete r C orpor- G-M detector One 9V battery X -rays 0-50 mR/hr and Two position range switch Used when a small compact s

ation M int-Rad and gamma 0-5 R/hr and ON-OFF battery switch device is desired l Radiation Monitor + 15%

Eberline Rascal Nine-inch diameter Five standard Neutrons 6 decades of rotary switch 0FF, H.V. ;. Used as a ratemeter or as i Mcdel PRS-2P cadnlum loaded poly- "D" size cells digital infor- rates A, B, C, and D;. a scaler for counting

{ ethylene sphere with nation minutes .5,1, 2, and 5 ; neutrons j BF3 tube in the manual, and stop. Two i center discriminator potentio-1 meters. Reset, speaker, scale illuminate and j gross /PHA switches.

i i

i l

i I

2563Q:1

i i TABLE 3-2 O

COMPARISON OF ESTIMATED OCCUPATIONAL DOSES FOR STEAM l GENERATOR DISPOSAL ALTERNATIVES l (MAN-REM)

O OPTION NUREG/CR-1595 m  :

i i

i  !

Long-term storage with 50 ,

cut-up and shipment l Long-tem storage with 30 I j intact shipment l Shorter-term storage l

, i

{ with cut-up - at 5 yr. 690 l - at 15 yr. 1 80 L

Immediate cut-up and 1700 shipment by rail / truck - no  !

i decontamination f i  !

I I

1 Immediate cut-up and 81 0 shipment by rail / truck - with chemical decontamination Immediate intact shipment 7 I O

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i SECTION l REACTOR COOL ANT PIPING i HOT t EG i

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Figure 3-6 Feedwater and Steam Line Piping Cut Points

4.0 RETURN-TO-SERVICE TESTING Following completion of steam generator repair, a preoperational test-N ing program will be conducted. This testing program will include the .-

following tests and inspections:

1. Reactor coolant system hydrostatic test

2. Secondary side hydrostatic test

3. Thermal expansion test during heatup l

4. Steam generator carryover test

5. Steam generator themal perfomance test l

6. Reactor coolant flow verification

7. Eddy current inspection of steam generator tubes  ;

e 8. Calibration and testing of instrumentation and controls affected by the repair activities

i These tests and inspections will provide the necessary assurance that  ;

the unit can be operated in accordance with design requirements and will j not endanger the health and safety of the public.

i O

i t

I 2516Q: 1 4-1 r----- ,- - + - - - - - - - - - - - , - - ----.-----.,r- - . - - - . - - - - - - - - - . + - - --r------ ------- - - - - .----e - ewm-

5.0 SAFETY EVALUATION O 5.1 FSAR EVALUATIONS 5.1.1 Introduction The purpose of this section is to evaluate the impact, if any, of the repaired steam generators on the accident analysis transients for Pcint i Beach Unit 1. Under the guidelines specified in 10 CFR 50.59 such an evaluation is required to verify that no unreviewed safety concerns or i changes to the Technical Specifications occur. This section provides a qualitative discussion of the effect on the accident analysis of steam generator parameter changes resulting~ from steam generator repair.  ;

Conclusions are made concerning the applicability of the original FSAR f to the repaired unit. Consistent with the requirements of 10 CFR 50.59, l licensing regulations and guidelines of the original licensing of the f Point Beach Unit are assumed to apply, and only changes in the safety analysis due to equipment changes are considered, i

The relevant plant operating parameters and steam generator design para-meters are compared in Table 5-1 and Table 5-2, respectively, for the i f original and repaired steam generators. While design improvements have l been incorporated, the repaired steam generators continue to match the design performance of the original steam generators. It may be noted from Tables 6-1 and 5-2 that there is very little change in original l plant operating parameters as a result of steam generator repair. It l is, therefore, to be anticipated that the impact on the accident [

analyses will be insignificant. The results of the accident evaluation  !

show that the repair of steam generators resulting in physically and  !

functionally similar units has not resulted in any adverse changes in l

the plant operating conditions used in the FSAR, and, therefore, the analyses presented in the FSAR are still valid. This section esta-  !

blishes that no unreviewed safety concerns exist due to operation with f

the repaired Point Beach steam generators. l i

O  !

i 2560Q: 1 5-1 l k

_ . _ _ _ _ , . _ _ - _ _ _ _ . . _ _ _ __ _._._____-___._.________-________t

i 5.1.2 Non-LOCA Accidents f

The Point Beach FSAR includes analyses of fifteen non-LOCA accidents in Sections 14.1 and 14.2. These are: *

a. Uncontrolled rod cluster control assembly (RCCA) withdrawal from a subcritical condition

b. Uncontrolled RCCA withdrawal at power f

c. Malpositioning of part length rods i

d. RCCA drop l

e. RCCA ejection i

f. Loss of reactor coolant flow I

g. Excessive load increase incident O h. Chemical and volume control system malfunction j

i

1. Startup of an inactive reactor coolant loop l

[

j. Reduction in feedwater enthalpy incident [

l

k. Loss of external electrical load

1. Loss of nonnal feedwater

m. Loss of all ac power to the station auxiliaries (blackout)  !

i

n. Liklihood of turbine-generator unit overspeed i

O  !

l 2560Q: 1 5-2  !

f I

I

, j r

o. Rupture of a main steam pipe (steam break)

The impact of the secondary system on the results of transients (a) through (e) is of no consequence since constant heat extraction is still maintained with the repaired steam generators. The main reason for this is that the nuclear and thermal time constants of the fuel are much O

v smaller than the fluid mixing and transport time, the latter being mechanisms responsible for secondary to primary interaction. For the ,

rod withdrawal and rod ejection accidents reactor trip occurs at a time near the magnitude of the coolant loop transport time. The limiting O consideration for the rod drop accident and malpositioning of part length rods is the neutron flux redistribution resulting from the con-trol rod movement and is clearly not coupled to steam generator perfor-mance. It can be validly concluded, therefore, that the first five accident transients named above are unaffected by the repair of the Point Beach steam generators.

Loss of reactor flow transients can be discussed collectively as was the case for the five reactivity insertion accidents above. Included in this general category are the following:

a. Total loss of reactor coolant flow

b. Partial loss of reactor coolant flow

c. Locked rotor If the reactor is at power, the immediate effect of loss of coolant flow is a rapid increase in coolant temperature, which could result in depar-ture from nucleate boiling (DNB). However, the reactor is tripped on low frequency, low voltage or low coolant flow trips such that the con-sequences are within the bounds of the FSAR analyses. The low-flow protection system, consisting of the low voltage, low frequency and low flow trips, rapidly detects and protects against loss of coolant flow O

2560Q:1 5-3

events. Changes in coolant temperature due to secondary parameter changes would not be detected in the core during the time frame of interest to this transient. Therefore, it can be concluded that the repair of the steam generator will not affect the loss of coolant flow transients.

The chemical and volume control system malfunction is a boron dilution in the reactor coolant system caused by adding unborated water to the reactor coolant via the makeup control system. Factors to be considered in the analysis of this transient are that the maximum dilution rate depends on the charging pump characteristics and that the malfunction i must be recognized and terminated by operator action. Repair of the steam generators with physically and functionally similar units will neither affect the initiating circumstances nor the corrective actions for the chemical and volume control system malfunction.

An excessive load increase incident is defined as a rapid increase in steam generator steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. The accident is analyzed in the FSAR to show that a 10 pen:ent increase in steam flow f from full power can be accommodated with a reactor trip. If a 10 per-cent step load increase is postulated, feed flow will increase to match l steam flow and maintain steam generator level. Depending on whether or '

not automatic rod control is available, a new steady state condition is l

established at the initial coolant average temperature or at a lower coolant average temperature. As is evident from the over-temperature AT  ;

equation, more than 10 pen:ent power margin in DNB is available. Repair of the steam generators resulting in units of similar physical size and tube structure cou1d slightly affect the excessive load increase acci-

~

dent in that the higher (by about 2 percent) full power fluid inventory of the repaired steam generators could cause the transient to progress more slowly; however, the same endpoint equilibrium condition would still be eventually reached, since no reactor trips are encountered.

O 5-4 2560Q: 1 t

The turbine generator design analysis describes the turbine generator and its speed control and provides a discussion concerning the velocity and energy of postulated ejected parts from the turbine. This analysis is completely independent of the nuclear steam supply system and thus is not affected by the repaired steam generators. i O It is apparent, therefore, that only those accidents which involve a primary-secondary interaction could potentially be affected by steam generator repair. Since the remaining accidents on the above list are generally concerned with primary coolant heatup or cooldown resulting from loss of secondary heat sink or excessive heat removal from the secondary side, they could potentially be affected by changes resulting from steam generator repair. These accidents are evaluated individually in the following sections.

5.1.2.1 Startup of an Inactive Reactor Coolant Loop 4

The cold leg temperature in the inactive loop will be identical to the  !

cold leg temperature of the active loops and to the reactor core inlet temperature. If the reactor is operated at power, there is a tempera-ture drop across the steam generator in the inactive loop, and reverse flow would exist through the inactive loop thereby lowering the hot leg  !

temperature of that loop below core inlet temperature. Administrative procedures require that the plant be brought to less than 10 percent load level and approximate temperature equilibrium between loops prior l to starting the pump in the inactive loop.

l p The startup of an inactive reactor coolant loop accident occurs when a '

V coolant pump in a loop, which contains water at a lower temperature than active loops, is started, causing a significant increase in water flow ,

into the core. The decrease in core temperature due to the increase in flow and the injection of colder water causes a rapid core power O' increase due primarily to moderator reactivity feedback. Verification that safety criteria for this accident are not violated is accomplished by demonstrating that the DNB ratio is always greater than 1.30.

O i

2560Q: 1 5-5

- 4 W The analysis presented in the Point Beach FSAR assined that the inactive p loop flow reversed and accelerated to its nominal full flow value instantaneously . The reactor coolant in that part of the inactive loop from the steam generator plenum to the reactor exit plenum (nonnally the hot leg) was assumed to be at a temperature equal to the saturation temperature of the secondary side. This assumption i s i ndependent of O the heat transfer characteristics of the steam generator, and thus is not affected by the repaired steam generators. Also, since the primary side volume is essentially unchanged for the repaired steam generators, the duration of the cold water slug '14 seconds) and the delay for the slug to reach the core inlet (7 seconds) would remain unchanged from the FSAR analysis. Therefore, the transient results presented in the FSAR would not be affected, and the accident criteria would continue to be met with the repaired steam generators.

5.1.2.2 Reduction in Feedwater Enthalpy Incident ,

The reduction in feedwater enthalpy is another means of increasing core p'ower above full power. Such increases are attenuated by the thermal c6pacity in the secondary plant and in the reactor coolant system. The overpower-overtemperature protection (nuclear overpower and AT trips) prevents any power increase which could lead to a DNBR less than 1.30.

An extreme example of excess heat removal by the feedwater system is the ,

transient associated with the accidental opening of the feedwater bypass valve which diverts flow around the low pressure feedwater heaters.

The function of the bypass valve is to maintain net positive suction  :'

A head on the main feedwater pump in the event heater drain pump flow is Q lost, e.g., during a lanje sudden load decrease. In the event of an accidental opening of the feedwater bypass valve, flow is diverted around the low pressure feedwater heaters. This causes a sudden reduc-

) tion in inlet feedwater temperature to the steam generators. Thi s increased subcooling will create a greater load demand on the primary system which can lead to a reactor trip.

2560Q: 1 5-6

Two cases are analyzed to demonstrate the unit behavior in the event of w a sudden feedwater temperature reduction resulting from accidental open-  ;

ing of the feedwater bypass valve. The first case is for the reactor in manual control with a zero moderator coefficient since this represents a condtion where the unit has the least inherent transient capability.

g The second case is for the reactor in automatic control with a large

, negative moderator coefficient. Initial pressure coolant temperature dnd power conditions are assumed at extreme values consistent with steady state operation to allow for calibration and instrument errors.

This results ia minimum margin to core DNB limit at the start of the transient.

During the accidental opening of a feedwater bypass valve transient, the secondary heat extraction is greater than the core power generation.

This causes the pressurizer pressure and coolant average temperature to decrease. Without automatic reactor control and a zero moderator coef-ficient of reactivity, the core power level increases slowly and eventu- .

ally comes to equilibrium at a slightly higher power level. With auto-matic reactor control and a 1arge negative moderator coefficient the negative coefficient causes the core power to increase rapidly. Steady state conditions are reached at a higher power level.

The analysis presented in the Point Beach FSAR for the accidental open-ing of a feedwater bypa:s valve without reactor control and zero modera-tor coefficient shows that Ta vg will decrease as secondary heat extraction remains greater than core power generation. As Tavg c o n- -

tinues to decrease the pressurizer heaters will not be able to maintain t pressurizer pressure and the reactor will be tripped by low pressure -

trip and the DNB ratio increases from the initial value. For the case with automatic red control and negative moderator coefficient the FSAR analysis shows that no reactor trip is generated and the DNB ratio remains well above 1.30. The replacer ent steam generator design has a larger full load steam generator mass, approximately 2 percent larger.

This increased secondary side heat capacity would result in a slightly lower cooldown rate than in the FSAR analysis. The steady state 2560Q: 1 5-7

conditions of the FSAR analysis would be reached at a slower rate. The same margins to reactor trip will exist. Therefore,.the accident criteria would still be met with the repaired steam generators.

5.1.2.3 Loss of External Electrical Load h

v A loss of external electrcal load may result from:

a. Abnormal variation in network frequency or other adverse network operating conditions

b. Trip of the generator or opening of the main breaker from the generator with failure of turbine trip. In this case the action of the turbine control system causes a large nuclear steam sup-ply system load reduction

c. Trip of the turbine ,

The unit is designed to accept a step loss of load from 100 percent to 8 50 percent load without actuating a reactor trip. The automatic steam bypass system, with 40 percent steam-dump capacity to the condenser, is able to accomodate this load rejection by reducing the severity of the

! transient imposed upon the reactor coolant system. The reactor power is reduced to the new equilibrium power level at a rate consistent with the i capability of the rod control system. Should the reactor suffer a loss of load from greater than 50 percent power, the reactor protection system would actuate a reactor trip.

l In the event the turbine bypass valves fail to open following a large load loss or in the event of a complete loss of load with steam dump operating, the steam generator safety valves may lift. The reactor I

(9 coolant temperature will increase rapidly and the DNB limit will be approached. The reactor is tripped on the folicwing signals:

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2560Q:1 5-8 l

a. High pressurizer pressure signal p

Q b. High pressurizer level signal

c. Overtemperature aT signal The steam generator shell side pressure will increase rapidly. The pressurizer safety valves and steam generator safety valves are, how-ever, sized to protect the reactor coolant system and steam generator against overpressure for all load losses without assuming the availabil-(O ity of the turbine bypass system. The steam dump valves will not be V

opened for load reductions of 10 percent or less. For larger load reductions, they may be open.

The most likely source of a complete loss of 1oad on the nuclear steam supply system i s a trip of the turbine-generator. In this case, there is a direct reactor trip signal (unless below approximately 10 pert:ent power) derived from either the turbine autostop oil pressure or a closure of the turbine stop valves. Reactor coolant temperatures and pressure do not significantly increase if the turbine bypass system and pressurizer pressure control system are functioning properly. However, in the Point Beach FSAR the behavior of the unit is analyzed for a com-l plete loss of load from 102 percent of full power without a direct j reactor trip due to a turbine trip primarily to show the adequacy of the 1

pressure relieving devices and also to show that no core damage occurs.

The reactor coolant system and main steam system pressure relieving capacities are designed to ensure safety of the unit without requiring the automatic rod control, pressurizer pressure control and/or steam bypass control systems.

In the Point Beach FSAR, the following cases are analyzed for the loss of external electrical load accident:

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a. BOL with pressure control and automatic rod control nv 2560Q: 1 5-9

b. E0L with pressure control and automatic rod control O c. BOL without reactor control and pressure control

d. EOL without reactor control and pressure control It is shown in the FSAR that the accident criteria on system pressure and DNB are not violated in any of the loss of load cases.

The slight increase in full load mass of the repaired steam generators would provide additional heat capacity and reduce the heatup rate. Thus the conclusions of the FSAR remain valid.

5.1.2.4 Loss of Nomal Feedwater A loss of normal feed water (from a pipe break, pump failure, valve malfunctions) could conceivably result in a loss in capability of the secondary system to remove the heat generated in the reactor core.

Since the plant is tripped well before the steam generator heat transfer capability is reduced and auxiliary feedwater flow initiated, the pri-mary system variables never approach a DNB condition.

The following provide the necessary protection against a loss of normal feedwater:

1) Reactor trip on very low water level in either steam generator

2) Reactor trip on main steam flow-feedwater flow mismatch in coincidence with low water level in either steam generator

3) Two motor driven auxiliary feedwater pumps (200 gpm each) which are started on O a. Low-low level in either steam generator O

5-10 2560Q:1

b. Opening of both feedwater pump circuit breakers

c. Any Safety Injection signal

d. Manually i

4) One turbine driven auxiliary feedwater pump (400 gpm) which is started on:

a. Low-low level in both steam generators or

b. Loss of voltage on both 4 kV busses, or

c. Manually On the loss of normal feedwater transient following the reactor and turbine trip, the water level in the steam generators will fall due to  !

the reduction of steam generator void fraction and because steam flow l through the safety valves continues to dissipate the stored and gen-erated heat. Making conservative assumptions on steam generator water  ;

level at time of trip, residual heat generation in core, number of ,

auxiliary feedwater pumps available and reactor coolant flow (natural l

{ circulation flow is assumed), the FSAR analysis demonstrates the ade- l quacy of the auxiliary feedwater system to remove stored and residual f heat without water relief from the primary system.

l The loss of normal feedwater is a loss of heat sink accident. The increased steam generator mass at full load of the repaired steam gen-erators is a change in a favorable direction. The physical dimensions of the steam generator have not changed. Therefore, the conclusion that the tubesheet in the steam generators receiving auxiliary feedwater will  ;

always be covered and adequate heat transfer capability will be main-tained remains valid.

l 2560Q: 1 5-11 i i

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5.1.2.5 Loss of All AC Power to the Station Auxiliaries (Blackout)

The loss of ac power to the station auxiliaries is analyzed to demon-strate long term heat removal capability by auxiliary feedwater and natural cirtulation reactor coolant flow.

In the event of a complete loss of offsite power and a turbine trip there will be a loss of power to the station auxiliaries, i.e., the reactor coolant pumps, main feedwater pumps, etc. Af ter a loss of ac power with turbine and reactor trip, the following events would occur:

a. Plant vital instruments are supplied by the emergency power sourtes.

b. Reactor coolant flow would coast down to natural circulation flowrate s. Main feedwater flow would stop and auxiliary feed pumps would automatically start.

c. The rise in steam system pressure following the trip would auto-matically open the steam system power operated relief valves. f (If the condenser is not available, the steam will be vented to the atmosphere.) If the steam flow rate through the power relief valves is not adequate, the steam generator self-actuated safety valves would lift to dissipate the sensible heat of the ,.

fuel and coolant above no-load temperature plus the residual l heat produced in the reactor.  ;

d. As the no load temperature is approached, the steam system power f relief valves are used to dissipate the residual heat and to l maintain the plant at the hot shutdown condition.

i Upon the loss of power to the reactor coolant pumps, coolant flow neces-sary for core cooling and the removal of residual heat is maintained by natural cirtulation in the reactor coolant loops. Differences in the average tube height could produce a change in the steam generator O i 2560Q:1 5-12 ,

contribution to the natural cin:ulation buoyancy driving head, and changes in the tube structure could cause a difference in primary steam system pressure drop. In the repaired steam generator design, the average tube height has not changed. Therefore, it is concluded that the loss of ac power analysis in the Point Beach is still applicable for the repaired steam generators.

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5.1.2.6 Rupture of a Main Steam Pipe (Steam Break)

A rupture of a steam pipe is assumed to include any accident which results in an uncontrolled steam release from a steam generator. An uncontrolled steam release, typically through a ruptured steam line or a defective valve, causes the secondary system temperature and pressure to fall and the heat transfer rate through the steam generator tubes to ri se . Therefore, the heat removal rate from the reactor coolant system increases, and the core moderator temperature drops. As the core is cooled, the neg tive moderator temperature coefficient causes the core reactivity level to rise.

The FSAR analysis of an uncontrolled steam release was performed to demonstrate that-

a. Assuming a stuck control rod assambly with or without offsite power and assuming a single failure in the engineered safety features, there is no consequential damage to the primary system .

and the core remains in place and intact.

b. There is no return to criticality for any single active failure in the main steam system. The single active failure is the opening, with failure to close, of the largest of any single steam bypass, relief, or safety valve.

c. Energy release to the containment from the worst steamline break does not cause failure of the containment structure..

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5-13 2560Q: 1

- - - - - - - --.. =. . -- - - .---

l The following systems provide the necessary protection against the steam f l break accident:  ;

! t

a. Safety injection system actuation on: ' !

1

. t l 1. One out of three pressurizer coincident low pressure and low I l

'l signals f l

l

2. Two out of three low pressure signals in any steamline  ;

i l 3. Two out of three high containment pressure signals

! b. Reactor trip on: i

! 1. Overpower reactor trips  !

i i

! 2. Reactor trip on safety injection signal .

(

, c. Feedwater isolation on safety injection signal l l d. Steam line isolation on:  ;

i i i

i

! 1. High steam line flow coincident with any safety injection l t l l signal or low reactor coolant average temperature j

! l l 2. Two out of three containment pressure signals  !

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i Major assumptions include use of end-of-life core kinetics parameters, i assumption of the most reactive rod cluster control assembly stuck in l

l the fully withdrawn position, and minimum safety injection capability j

! due to a single failure in the system. f

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, i j

The cases considered are the complete severance of a main steam pipe l upstream and downstream of the flow restrictor in the steam pipe, with  ;

i and without the simultaneous loss of offsite power and steam release i

O l i

i j 2560Q: 1 5-14 f u _ _ _ _._ . _ _ _ _ . _ . _ . _ _ . _ _ _ _ . _

i l

through a safety valve. All the cases assume initial hot shutdown con-dition since steam generator mass inventory is greatest at that condi-tion. Should the reactor be just critical or at power at the time of the steam line break the reactor would be tripped by the nonnal over-power protection system and the additional stored energy would be removed by the cooldown before the no load condition and shutdown margin assumed above are reached. In addition, the greater steam generator mass at hot shutdown conditions increases the magnitude and duration of l the cooldown.

l O The core power and reactor coolant system transients will not be affected by the repaired steam generators. The reasons for this con-clusion include the following: i

a. The key parameters which strongly influence the transient are performance of the emergency shutdown system and core reactivity coefficients. There are no changes to these parameters as they j are used in the analysis due to repair of the steam generators.

b. The flow area of the main steam line is an important factor in  !

detennining the amount and rate of heat extracted from the i reactor coolant. This flow area decreases due to the integral flow restrictors. The FSAR analysis would be bounding, since no

. credit was taken for intergal restrictors. >

4 l

c. No changes are expected due to differences in initial conditions j (zero load steam temperature and pressure are identical for the unit with repaired steam generators). The no load steam genera-O tor mass increases insignificantly (~1.5 pen:ent).

Therefore the steam line break analyses presented in the FSAR are valid for the repaired steam generators.

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25600: 1 5-15

e 5.1.3 Loss of Coolant Accident Evaluation d In the event of a major reactor coolant system pipe break, depressuriza-tion of the reactor coolant system results in a pressure decrease in the presturizer. Reactor trip signal occurs wnen tne pressurizer low pres-sure trip setpoint is reached. A safety injection system signal is l O

\ actuated when the appropriate setpoint is reached. These counter-measures will limit the consequences of the accident in two ways:

a. Reactor trip and borated water injection complement void forma-i tion in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

l b. Injection of Dorated water provides neat transfer from tne core

! and prevents excessive clad temperatures.

At the beginning of tne blowdown pnase, the entire reactor coolant sys-tem contains subcooled liquid which transfers heat from the core by i forced convection with some fully developed nucleate boiling. After the break develops, the time to departure from nucleate boiling is calcula-ted, consistent with Appendix K of 10 CFR 50. Thereaf ter, the core heat l

transfer is based on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms. During the

, refill period, rod-to-rod radiation is the only heat transfer mechanism. .

l Wnen the reactor coolant system pressure f alls below 600 psia, the accu-mulators begin to inject borated water. The conservative assumption is made that accumulator water injected Dypasses the core and goes out through the break. This conservatism is again consistent with Appendix K of 10 CFR 50.

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2560Q:1 6-16 i

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The reactor is designed to withstand themal effects caused by a loss of coolant accident, including the double ended severance of the largest reactor coolant system pipe. The reactor core and internals together with the emergency core cooling system (ECCS) are designed so that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident.

The ECCS, even when operating during the injection mode with the most severe single active failure, is designed to meet the acceptance cri-teria (Reference 1).

o Several large break loss of coolant analyses have been submitted as amendments to the Point Beach FSAR. The most applicable existing analysis to use as a baseline for the evaluation of the steam generator repairs was perfomed with 18 percent steam generator tube plugging and l Fq equal to 2.32.

This analysis utilized the Westinghouse, February 1978 ECCS Evaluation Model (References 7, 8, 9 and 10)).

The analysis was perfomed with a reactor vessel upper head fluid tem-perature equal to the reactor coolant system hot leg temperature. The effect of using the hot leg temperature in the reactor vessel upper head is described in Reference 6.

An evaluation was perfomed to detemine the effects on ECCS perfomance of the repaired steam generator parameters. The evaluation basis was a i comparison with the aforementioned large break analysis for Point Beach  ;

> which incorporates 18 percent steam generator tube plugging with an F q of 2.32.

i i

O 2560Q:1 5-17 i

The evaluation shows that the effect of 18 percent tube plugging is more limiting with respect to ECCS performance than the effects of the re-b paired steam generator parameters. Thus, the previous 18 percent tube plugging analysis is conservative for the Point Beach Plant with repaired steam generators.

O The slight decrease in primary side volume and heat transfer area will not impact the blowdown portion of the LOCA analysis since these changes are only second order effects during tnis phase of the accident.

O The tube resistance for the replacement steam generators will be lower than that of the original steam generators at the existing analysis conditions. Therefore, the impact on LOCA PCT will be a benefit over the previous analysis due to the improvement in core reflood rates.

From this evaluation it is concluded that the repaired steam generators, compared to the original steam generators, would snow an improvement in the limiting break LOCA analysis results.

The existing small break analysis for Point Beach is described in Sec-tion 14.3 of the FSAR. This analysis is in conformance witn 10 CFR 50.46 and Appendix K to 10 CFR 50. None of the parameters in the re-

' paired steam generator has a significant effect on small break LOCA.

Thus the effect on the small break analyses in negligible and the existing small break analyses in the FSAR are applicable to the plant with repaired steam generators.

l The containment mass and energy release in the FSAR were performed in conformance with criteria existing at the time of the Point Beach Oper-ating License submittal. This analysis for a LOCA is sensitive to re-

, actor power and primary system volume. Since both of these parameters I are virtually unchanged, the containment mass and energy release would not change and the existing analysis is still applicable.

l l

O 2560Q: 1 5-18 i

--w- - , , - - ,-mp 9-- y. p ,,p -

--Tr -e -**e-T"7 --aw g- r- w +- -e-m -*---- - -- --- - -m-y e==-r-- - - - - - w- rw--ew-My--- eum v

5.1.4 Steam Generator Tube Rupture O No plant parameters will change as a result of the steam generator re-pair, and, as shown in Tables S-1 and 5-2, no steam generator parameters will be changed which would affect the tune rupture analysis. Thus, the tube rupture analysis and consequences in the Point Beach FSAR would be O unchanged with the repaired steam generators and remain valid.

5.1.5 Refe:ences for Section 5.1 O 1. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50. Federal Register, Volume 39,. Numoer 3, January 4, 1974.

2. Bordelon, F.M., et al., " Westinghouse ECCS Evaluation Model -

Supplementary Infonnation," WCAP-8471, April 1975 (Proprietary) and WCAP-8472, April 1975 (Non-Proprietary).

3. " Westinghouse ECCS Evaluation Model October 1975 Version," WCAP-8622 November 1975 (Proprietary), and WCAP-8623, November 1975 (Non-Proprietary).

4. Letter from C. Eicheldinger of Westinghouse Electric Corporation to D. B. Vassallo of the Nuclear Regulatory Commission, Letter Number NS-CE-924, January 23, 1976.

i

5. " Supplement to tne Status Report oy the Directorate of Licensing in the matter of Westinghouse Electric Corporation ECCS Evaluation Model Conformance of 10 CFR 50 Appendix K, " Federal Register, November 1974.

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2560Q: 1 5-19

6. Letter from C. Eicheldinger of Westinghouse Electric Corporation to l V. Stello of the Nuclear Regulatory Commission, Letter Number O NS-CE-Il63, August 13, 1976.

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I i

7. "Westinpouse ECCS Evaluation Model, February 1978 Version", l

, WCAP-9220-P-A (Proprietary Version)), WCAP-9221-A (Non-Proprietary  ;

Version), February,1978. ,

8. Letter from T. M. Anderson of Westinghouse Electric Corporation to l J. l

, John Stolz of the Nuclear Regulatory Commission, Letter Number 4 NS-TMA-1981, November 1, 1981. l 1

9. Letter from T. M. Anderson of Westinghouse Electric Corporation to R. L. Tedesco of the Nuclear Regulatory Commission, Letter Number NS-TMA-2014, December 11, 1978.

i

10. " Safety Evaluation Report on ECCS Evaluation Model for Westinghouse l Two-Loop Plants", Novemoer,1977.

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2560Q

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5.2 CONSTRUCTION RELATED EVALUATIONS Rigging and transportation of heavy load requirements have been evalu-

.%[ )] ated. Administrative procedures and controls will be established to minimize the potential for of any mishap. Nevertheless, the potential for rigging and construction incidents to occur have been postulated.

The evaluations below demonstrate that the in situ configuration,

^} augmented where appropriate with temporary physical protection, can accommodate all events analyzed.

f) The conclusion reached by the analysis of construction related incidents is that any loss of safety-related functions has been precluded. There-fore, there is no unreviewed safety question associated with the con-struction activity.

5.2.1 Handling of Heavy Objects As described in Section 3.1.5, precautions will be taken to preclude the possibility that a rigging or transportation incident will adversely 8 aifect any component, system or structure important to the nuclear safety of either unit. These precautions include training of equipment operating personnel, additional protection of buried piping and duct banks where necessary along the steam generator haul routes, controls on haul routes and equipment speed, and controls on lift heights, travel directions, location and swing arcs for both loaded and unloaded cranes. However, for the purpose of evaluation, certain construction related incidents were analyzed and the results of the analyses are summarized in the following subsections.

O'%J These analyses demonstrate that the spectrum of postulated events will not preclude the ability to achieve / maintain a safe shutdown condition.

These events are not likely to occur. In all cases analyzed, these events have been precluded by design and/or temporary augmented protec-tive measures.

b

'V 2560Q: 1 5-21

r l

5.2.1.1 Overturning of a Loaded Trailer i Analyses were performed to determine the conditions which would be required to overturn a typical trailer loaded with a steam generator lower assembly. The following assumptions were made for the purpose of tnese analyses: t

a. A multi-axle, multi-tire trailer with a bed height of 4 feet above the ground.

b. Trailer tire span width of 12 feet out-to-out of tires.

c. A combined trailer weight and steam generator saddle weight of l P

70 tons with a center of gravity at 4 feet from ground elevation. ,

d. A steam generator lower assembly weight of 205 tons with center of gravity at 12 feet from ground elevation. l

c. A worst case turn radius of 25 feet.

O Results of analyses indicate that the loaded trailer would become unstable and overturn if inclined beyond 31 degrees from the horizontal, l or if the trailer exceeds a speed of 15 miles per hour in a turn radius of 25 feet.

For the following reasons, it is concluded that overturning of a typical trailer is highly unlikely:

a. The haul routes do not have any banking; therefore, the trailer will not be inclined laterally. In addition, as described in

Subsection 3.1.1.2, haul routes will be evaluated and upgraded, l if necessary, to preclude the possibility of roadway collapse. j O r i

i l

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2560Q: 1 5-22

b. The turning radius of any haul road on the plant site will sig-nificantly exceed the conservatively assumed turn radius of 25 (v feet.

c. A speed less than 15 mph will be maintained by administrative control s.

O 5.2.1.2 Effect of Postulated Rigging Incidents en Safety-Related Plant l Structures Postulated Crane Boom Drop (Vn)

There are no safety related structures within range of a 360* radius of any crane boom in the working area adjacent to the Unit 1 containment hatc h.

Postulated Steam Generator Lower Assembly Drop _

There are not subterranean structures at the containment hatch either safety-related or otherwise. Therefore, in the event that a steam gene-rator is dropped while being transferred to or from the trailer, no affect on to an existing structure is expected.

l l

The structures which could be impacted by rigging incident have not been evaluated further because they do not perform safety-related functions during steam generator repair on that unit. These structures are as follows and related to the corresponding location identified in Fig. 5-1:

, 0) x (Location A) Containment - During the repair, fuel is.

removed from the affected l containment.

('

(Location B) Service Bldg. -

0 2560Q:1 5-23

(Location C) Extension Bldg. -

(Location D) Overhead Electrical Wires -

Three overhead electrical wires which pass over the work area and possibly j within range of the crane ,

boom (s) will be out of ser-vice during Unit 1 outage.

Although normal construction l

practices will be followed -

to prevent contact with the wires, a postulated accident i which results in severing  ;

any or all three lines will l have no consequential impact j on the safety of the plant l t

or the workers since the i lines will be out of service with the shutdown of Unit 1. I O

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2560Q: 1 5-24 l  !

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5.2.2 Radioactive Releases and Dose Assessment ,

Radioactive airtorne and liquid releases have been evaluated for the repair effort using conservative, bounding parameters and assumptions.

In order to assess the significance of these releases, they were compared with the radioactive releases at Point Beach in the year 1981.

The total calculated release per unit for the repair effort was found to be a small fraction of the 1981 radioactive releases per unit and, therefore, is acceptable.

5.2.2.1 Airborne Releases Airborne effluent releases to the environment resulting from this type of repair effort have been estimated in Reference 1 as follows:

The primary airborne releases of radionuclides auring steam generator removal are due to 1) cutting the reactor coolant piping and 2) cutting other system piping. Containment envelopes are assumed to be used when cutting the reactor coolant piping. These containment envelopes have a HEPA filter in their ventilation system and are exhausted through the plant ventilation system. For other cutting operations, no containment envelopes are assumed. For these calculations, it is assumed that all HEPA filters are preceded by a demister, which is necessary to retain filter integrity. Segmenting the steam generator at the transition cone and the internal wrapper does not contribute significantly to airborne releases because the contamination levels on the secondary side of the ges.erator are several orders of magnitude below those on the primary I side.

O Airborne releases were calculated as follows:

Cutting the reactor coolant piping O 1. Four cuts with a 0.95-cm kerf are made in 86-cm-ID pipe.

2 l 2. 4 x 0.95 x 86 w = 1030 cm of material vaporized.

O 2560Q: 1 5-25

3. The contamination level on the interior of the piping is 86 uC1/cm2 (see Table 5-3).

2

4. 1030 cm x 86 uCi/cm2 = 8.9 x 104 uCi released.

5. With a decontamination factor of 104 (two HEPA filters preceded by demisters), release to the atmosphere is 8.9 uCi per steam generator..

Cutting other system piping

1. Single cuts with a 0.95-cm kerf are made in six 15-cm-ID and six 5-cm-ID pipes, and two cuts are made in one 76-cm-ID (steam line) and one 36-cm-ID (feedwater) pipe.

2. 0.95 x (6 x 15 x + 6 x 5 x + 2 x 76 1 + 2 x 36 w) = 1.0 x 10 3 cm2 of material vaporized.

3. The contamination level on the interior of the pipes is 6.2 uCi/cm2 (see Table 5.2-1).

O 4. 3 2 1.0 x 10 cm x 6.2 uCi/cm2 = 6.4 x 103 Ci released.

5. With a decontamination factor of 102 (one HEPA filter preceded by a demister), release to the atmosphere is 64 gCi per steam generator.

Assuming two steam generators per reactor unit, the total release for cutting operations would be 146 uCi. The radionuclide distribution would be similar to that listed in Table 5-4.

O Actual airborne releases (Table 5-5) measured during the steam gener-ator replacement at Surry Unit 2 were 101.3 Ci of noble gases, 6.88 x 10-6 Ci of iodines, and 1.32 x 10-3 Ci of particulates.(2) Air-borne releases from fuel unloading and reloading are included in the Surry measurements.

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t 2560Q: 1 5-26

5.2.2.2 Environmental Consequences of Airborne Releases <

The critical organ and whole body doses for an adult at the worst site f

boundary location resulting from the estimated airborne effluent releases during the repair effort were evaluated using Reference 1 ,

assumptions, an annual average a ground level release atmospheric O

s dispersion factor of 1.5 x 10-6 sec/m ,3and the dose models and dose l

factors given in Regulatory Guide 1.109. The critical organ (lung) and  !

the whole body doses for an adult at the site boundary are estimated to  :

be 2.8 x 10-5 mrem and 6.8 x 10-8 mrem respectively during the repair effort for Unit 1. ,

5.2.2.3 Comparison witn Observed Gaseous Releases and Estimated Doses During Normal Operation i

The estimated releases of radioactive airborne effluents per unit during i t

the repair effort are found to be much smaller than the observed gaseous  ;

effluent releases per unit for the Point Beach Plant during the year l 1981. Observed gaseous effluent releases during 1981 are compared with .

estimated releases during the repair effort in Table 5-6.

i The critical organ (thyroid) and whole body doses to an adult at the worst site boundary location due to the release of gaseous effluents for [

the year 1981 were calculated to be 0.003 and 0.07 mrem / unit, respec-tively. The estimated critical organ (lung) dose for the repair effort is less than 1.0 percent of the calculated critical organ dose during 1981. The estimated whole body dose for the repair effort is less than 0.0001 percent of calculated whole body dose during 1981.

5.2.2.4 Liquid Effluent Releases Liquid effluent releases resulting from tne repair effort were estimated [

using the following parameters and assumptions:

a. The reactor coolant system is drained 15 days after reactor ,

shutdown and the reactor coolant is subsequently discharged l (about 30 days after shutdown) after processing through a mixed 2560Q:1 5-27 l - -- _ __ .- . _

bed demineralizer and through the boric acid recovery evaporator ,

as required. Laundry waste water is discharged without processing.

b. The decontamination factors for processing equipment are listed below and are in accordance with NRC NUREG-0017:

Processing Equipment Decontamination Factors r

Ioidines Cs and Rb Others Mixed bed demineralizer 10 2 10 Boric acid recovery evaporator 100 1,000 1,000

c. Reactor coolant concentrations are given in Table S-7 and are i based on values given in Reference 1 which were taken from NRC NUREG-0017. These values are conservative since Point Beach Unit i reactor coolant concentrations have been generally much lower.

i

d. The mass of tne reactor coolant discharged after processing is 5 '

2.6 x 10 lbs.

i

e. Laundry releases were estimated using the expected specific l activities in the laundry waste water given in Table 5-8 and '

assuming approximately 26,000 gal / day of laundry waste water will be discharged for approximately 180 days during the repair effort for one unit. (It is expected, however, that on the average only 10,000 gal / day of laundry waste water will be ,

discharged during this period.)

The total radioactive liquid effluent release based on the above l assumptions is estimated to be approximately 0.23 Ci/ unit (chiefly laundry waste), excluding tritium and dissolved gases, and approximately j 125 Ci/ unit of tritium. Details of this release by isotope are given in  !

Tables 5-9 and 5-10.

2560Q: 1 5-28

5.2.2.5 Comparison with Observed Radioactive Liquid Releases During Nomal Operation g

Estimated radioactive liquid releases during the repair effort are compared with the observed liquid waste releases during the year 1981 in Table 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 pen:ent of the observed total liquid waste release per j unit (excluding tritium and dissolved gases) during 1981. The estimated  !

tritium release per unit during the replacement effort is about 38 '

percent of the observed tritium release per unit during 1981.

5.2.3 References for Section 5.2  ;

1. Hoenes, G.R., M. A. Mueller, W. D. McComack,1980, Radiological Assessment of Steam Generator Removal and Replacement: Update and Revision, NUREG-CR-1596 (PNL-3454), U.S. NRC, Washington, D.C.

2. Virginia Electric and Power Company,1979, Stearn Generator Repair Program for the Surry Power Station Unit No. 2 - Final Report i (Progress Report - No. 6) for the Period February _3,1979 through December 31, 1979, NRC Docket Numbers 50-280 and 50-281, Washington, f 0.C.  !

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t TABLE 5-1 [

O COMPARIS0N OF OPERATING PARAMETERS FOR ORIGINAL ANO REPAIRED STEAM GENERATORS PARAMETER i

O

Nominal power /SG Unchanged i

i Nominal primary flow /SG Unchanged O [

Nominal hot leg temperature, *F Unchanged

[

4 Nominal cold leg temperature, 'F Unchanged l Feedwater temperature, *F Unchanged i i

b Reactor Coolant System pressure, }

i

psia -

Unchanged l O

Nominal steam pressure, psia Unchanged ,

I

! Nominal fluid mass /SG, lbm Increased f 2 percent  !

No load fluid mass /SG, lo" Increased 1.5 percent ,

i No load temperature, *F Unchanged i

i Steam flow Unchanged O l i i 1

j .

i 2560Q:1 l i

_ __._ _ .__ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ _ _ . . _ _ _ - _ ___.__-.._.____._J

4 TABLE 5-2 COMPARIS0N OF DESIGN PARAE TERS FOR ORIGINAL AND REPAIRED STEAM GENERATORS i l l

Primary Side Volume Decreased by less than .4 percent l Number of Tubes Decreased by 46 l 4

Tube 0.0. Unchanged Wall Thickness Unchanged I

4 l

Primary Pressure Drop Decreased by 0.3 psi Fouling Factor dnchanged Heat Transfer Area Decreased by 2.2 percent i . i Flow Area Decreased by 1.5 percent 1

l Equivalent Length Increased by 1.5 percent l

l l

O -

i, O i O

l 2560Q: 1

i 1

1 TABLE 5-3 i

Gross Contamination Levels by Location in Piping i and Steam Generator (Reference 1) I f i 1

Contamination Level, 2

i Component uCf/cm ,

I Reactor coolant piping 86 '

l Other piping 6.2 l

i i Steam generator 1 l

Primary side j tubes 8.2 k tubesheet 140 channel head 68 partition plate 140 i Secondary side ~10-3 i

I l

O l i

f l

O  !

e O  !

r

~~

2560Q: 1  !

i k

I i

l l l TABLE 5-4  !

l Point Beach Nuclear Plant, Unit 1 l Estimated Steam Generator Curie Content i  !

i Isotog Percent of Total Estimated Curies l

O 60 Co 62 186 fI i

l 58 Co 26 78 l

' O 103 Ru 5.8 17.4  !

l 141 Ce 3.2 9.6 1 l 51 Cr 2.9 8.7 i-l l l

Balance

• 0.1 0.3 TOTAL 100 300 1  !

f i

[.

r i

I i

O i

i O l

• 0ther isotopes identified include 95 Zr, 95 rio,137 Cs,144 Ce, t 241 Am, 109 Cd, 140 La, 59 Fe, 54 Mn, 124 Sb and 106 Ru. j O  !

l

~

2560Q: 1 -

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

r j TABLE 5-5 l Effluent Release Isotopic Distributions Steam Generator

• Replacement Project Surry Power Station - Unit No. 2  :

l i

(Reference 2)  ;

i GASEOUS EFFLUENTS l

Total Activity Percent Released of Total

] '

Isotope (Cf) Activity t

i Noble Gases

t 4

Xe-133 99.4 98  !

l Xe-135 1.9 2

u Total 101.3 100

[

l O Iodines i

I-131 5.88 x 10-6 100  !

I Total 6.88 x 10-6 100 l

I r

Particulates {

Co-60 7.00 x 10-4 53 l Co-58 3.01 x 10-4 23  !

Cs-137 2.19 x 10-4 16 {

Cs-134 4.94 x 10-5 4 l Cr-51 4.51 x 10- 3 i Mn-54 8.37 x 10-6 1 Total 1.32 x 10-3 100 2560Q:1  !

I

d i

! i a 1 i . t I TABLE 5-6 I i i Comparison of Gaseous Effluent Releases i f Average Estimated Release I j g 1981 During the SG

\J/ Release / Unit Repair Effort Isotope (C1) (Ci) l j Noble gases 305 Negligible  !

j t' Iodines 5.2 x 10-3 6.8 x 10-6(1)

I  !

l Particul ates 9.6 x 10-2 1.46 x 10-4 i

i Tritium 240 Negligible Notes f (1) Estimated from Surry Unit 2 Data, Table 5.2-3.

i i

O l i

O O l i

2560Q: 1 [

TABLE 5-7 Radionuclide Concentrations in Reactor Coolant (Reference 1)

Radio- Hal f-Li fe, Concentration Radio- Hal f-Li fe , Concentration O nuclide days uCi/g nuclide days uCi/g 3g 4.51E+03(a) 1.0E+00 106 Rh 3.46E-04 1.0E-05 16 8.22E-05 4.0E+01 125m 5.80E+01 2.9E-05 Te O 51 N

Cr 2.77E+01 1.9E-03 3.1E-04 127m Te 1.09E+02 3.90E-01 2.8E-04 8.5E-04 54 3.13E+02 127 Te Mn 557 , 9.86E+02 1.6E-03 129m 3.36E+01 1.4E-03 Te 59 4.46E+01 1.0E-03 129 Te 4.83E-02 1.6E-03 Fe 58 7.08E+01 1.6E-02 131mTe 12.5E+0 2.5E-03 Co 60 1.93E+03 2.0E-03 131 Te 1.74E-02 1.1E-03 Co 83 9.96E-02 4.8E-02 132 3.26E+00 2.7E-02 Br Te l

84 gp 2.21E-02 2.6E-03 130 7

5.'15E-01 2.1E-03 85 1.99E-03 3.0E-04 131 8.04E+0 2.7E-01 Br 7 86 1.87E+01 8.5E+01 132; 9.5E-02 1.0E-01 l Rb 88 1.24E-02 1.0E-01 133; 8.67E-01 3.8E-01 Rb 89 5.06E+01 3.5E-04 134 g 3.65E-02 4.7E-02 37 90 3p 1.04E+04 1.0E-05 135 g 2.75E-01 1.9E-01 91 3.96E-01 6.5E-04 134 7.53E+02 2.5E-02 37 Cs 90 y 1.67E+00 1.2E-06 136 Cs 131E+01 1.3E-02 91my 3.4E-02 3.6E-04 137 1.10E+-4 1.8E-02 Cs 91 y 5.81E+01 6.4E-05 137m 1.78E-03 1.6E-02 l Ba 93 y 4.21E-01 3.4E-05 140 1.28E+01 2.2E-04 Ba 95 6.40E+01 6.0E-05 140 1.68E+0 1.5E-04 Zr La 95 3.52E+01 5.0E-05 141 3.25E+01 7.0E-05 Nb Ce

~99gg 2.75E+0 8.4E-02 143 1.38E+0 4.0E-05 Ce 99m 2.51E-01 4.8E-02 144 2.84E+02 3.3E-05 Tc Ce 103 3.93E+01 4.5E-05 143 pp 1.36E+01 5.0E-05 Ru 106 3.68E+02 1.0E-05 144 pp 5.00E-03 3.3E-05 Ru 103mRh 3.90E-02 4.5E-05 239 3p 2.35E+00 1.2E-03 (a) 4.51E+03 = 4.51 x 10 3 2560Q: 1 l

TABLE 5-8 Estimated Specific Activities of Laundry Waste Water Specific Activity III Isotope u Ci/cc Co-58 6.7 x 10-6 Co-60 5 x 10-6 O Cs-137 5.4 x 10-6 Cs-134 6.5 x 10-7 Mn-54 7.3 x 10-7 I-131 1.1 x 10-7 O Note (1) Time-averaged specific activity during a period of 180 days.

O O 1 l

1 1

0 25600:1 i

TABLE 5-9 Estimated Radionuclude Releases Due to Discharge of Reactor Coolant Water (a)

Radionuclide Release, Ci O 3H 1.25E+02(b) 51C r 1.1E-05 54Mn 3.6E-06 55Fe 2.0E-05 59pe 7.9E-06 O

I 58Co 1.5E-04 i 60Co 2.5E-05 l 1.7E-06 86Rb 89S r 2.9E-06 90S r 1.3E-07 90y 1.3E-07 91y 5.6E-07 95Z r 5.4E-07 95Nb 3.5E-07 99go 5.5E-07 3.3E-07 O 103Ru 106Ru 125mie 1.2E-07 2.5E-07 127mTe 2.9E-06 129mTe 9.2E-06 131mTe 1.8E-12 132Te 5.7E-09 131I 2.6E-03 1331 1.8E-12 134Cs 1.5E-03 136Cs 1.6E-04 137Cs 1.1E-03 O 140Ba 140La 141Ce 5.4E-07 7.9E-12 4.6E-07 143Ce 1.4E-13 144Ce 3.8E-07 O 143P r 239 Np 1.3E-07 2.2E-09 Total 1.25E + 0.2 including tritium 5.6E - 0.3 excluding tritium (a) For a power plant with two steam generators O (b) 1.25E+02 = 1.25 x 102 2560Q:1

4 3

TABLE 5-10 Estimated Radioactive Liquid Effluent Releases During the Steam Generator Repair i

..t i

Release / Unit f Isotope (Ci)  !

l Mn-54 1.1 x 10-2 l i

Co-58 1.0 x 10-1 4

i Co-60 7.5 x 10-2 i

j Cs-134 1.1 x 10-2 l l

Cs-137 3.2 x 10-2  ;

I-131 4.3 x 10-3 lO i l

l Total 2.3 x 10-1 l l Tritium 125 i

i i

I f

O  !

I l

l l

I 2560Q: 1 ,

. ____________._..____..___._.F

TABLE 5-11 ,

Comparison of Radinactive Liquid Effluent Releases l 1

m erage i

1981 Estimated Release During the S.G.

O i

Release / Unit Repair EffortI II  !

Isotope (Ci) (Ci) j i

' i Total 0.55 0.23  !

(excluding tritium and 5 dissolved gases) i J

h Tritium 326 125 i

, (1) The total releases excluding tritium, estimated for steam generator  !

repair activities, conservatively assumes no processing of laundry I wastes prior to release. It is expected that these wastes will be f j processed in the plant radioactive waste processing system. Such  !

processing would reduce the estimated releases by at least an order -

cf magnitude. <

i i

I i

i O

i l

f 2560Q:1  !

?

6.0 ALARA CONSIDERATIONS p 6.1 ALARA OBJECTIVES O

The steam generator replacement activities described herein will be implemented at the Point Beach Nuclear Plant No.1 which will have oper-p ated for approximately thirteen 13 years at the time of repair. As a b result of irradiation and contamination, many of the tasks associated with the replacement activities will expose personnel to radiation.

Radiation exposure to personnel will be maintained at levels that comply p with 10 CFR 20, " Standards for Protection Against Radiation". In addi-O tion, every reasonable effort will be made to maintain exposures to radiation below the limits specified in 10 CFR 20 and as low as is reasonable achievable. Accordingly, this section provides infomation relevant to attaining goals and objectives for planning, designing, engineering, and implementing the steam generator replacement activities to assure that exposures of personnel will be "as low as is reasonably achievable" (ALARA). The guidance and recommendations contained in Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupa-

' tional Radiation Expsoures at Nuclear Power Stations Will Be As Low As Is Reasonabley Achievable", Revision 3, June 1978 have been considered j

in fomulating the ALARA Program for the project.

The overall goal of the ALARA effort for the replacement activities are (1) to maintain the annual dose to individual personnel as 1ow as reasonably achievable and (2) to keep the annual collective dose to personnel, i.e. total man-rem exposure, as low as reasonably achievable.

In satisfying these objectives, both current technology and good work l V practices will be used. The means available to implement ALARA ob-jectives are varied and must be applied on a case-by-case basis. In j

evaluating ALARA considerations, some decisions are obvious without a detailed study, whereas cost benefit analyses are used in some cases.

J The specific objectives which have been established for the replacement activities to maintain occupational radiation exposures ALARA are:

' h o

2614Q: 1 6-1

A. Establish that the ALARA philosophy shall apply to the program O B. Engineer facilities, systems and components, and select equipment, that satisfy ALARA objectives C. Establishment of a radiation control program, plans and procedures

6. Make supporting equipment, instrumentation and facilities available E. Preplan and schedule all activities to minimize duration, exposure and number of people F. Assign trained personnel to perform replacement activities 6.2 STEAM GENERATOR ALARA PROGRAM For purposes of discussion, the information presented herein will be in '

the context of the steam generator replacement activities, although many '

of the principles.and practices identified m'ay be applicable for normal plant operation, and have been applied to those activities. Some of the items identified are unique to tne replacement program. For steam gen-erater repair, the ALARA Program has not been developed as an inde-pendent document, but is included as part of Inis suomittal and other l special instructions or procedures issued for the steam generator project. .

As stated herein, there is a management policy for, and commitment to, ,

ensuring that the exposure of personnel to radiation is ALARA. This policy and commitment is implemented by existing procedures and in-structions and by special procedures applicable to the replacement I

activities. The activities involved with the overall replacement are identified as discrete " work packages". These detailed work packages O

Q include appropriate procedures, instructions, drawings, etc. to assure that the work can be c.xnpleted with a minimum of radiation exposure.  !

Tnese work packages include any special provisions that are necessary in O

! 2614Q: 1 6-2 1

performing the work, e.g. temporary shielding. Special emphasis has been placed on the engineering and planning prior to issuing " work pack- -

G ages" to assure that ALARA objectives are incorporated.

The steam generator project organization reflects a commitment to ALARA objectives. A full time engineer knowledgeable with health physics practices is assigned to the headquarters organization to assist in the planning phase. During the engineering and planning phase, ne is assisting in establishing basic criteria for implementing ALARA objec-tives, as well as working with tne engineers, planners and consultants O in establishing specific requirements. During the actual work a full )

time Health Physics Director, knowledgeable with healtn physics prac-tice, will implement all health physics activities, through Health Physics Shif t Coordinators and Health Pnysics Technicians. The Health Physics Director will have line responsibility through the Site Manager and Program Manager to the WE Special Projects Administrator.

Health physics procedures written for the steam generator project will be utilized. These procedures provide instructions for HP related items 9 and implement applicable Point Beach Health Physics Procedures and Policies. The Health Physics Director will oe responsible for assuring 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 De com-promised. The appropriate resources needed to acnieve ALARA goals and objectives will be provided. The Health Physics Director and Health Physics Shift Coordinators will be able to call upon the headquarters organization as well as consultants for support.

The basic responsibilities of the Health Pnysics Director for tne replacement activities includes:

A. Participating in design reviews for f acilities and equipment tnat can affect potential radiation exposures; O

2614Q:1 6-3

B. Identifying locations, operations, and conditions that hav.e the potential for causing significant exposures to radiation; C. Initiating and implementing an exposure control program; O. Developing plans, procedures, and methods for keeping radiation exposures of plant personnel ALARA; E. Reviewing, commenting on, and recommending changes in job procedures to maintain exposures ALARA; F. Developins and participating in training programs related to work in radiation areas or involving radioactive material; '

G. Supervising the radiation surveillance program to maintain data on exposures of and doses to plant personnel, by specific job functions and type of work; H. Supervising the collection, analysis, and evaluation of data and i information attained from radiological surveys and monitoring activ-ities; l

l I. Ensuring that adequate radiation protection coverage is provided for plant personnel during all working hours.

l 1

J. Coordinate activities of the steam generator project with those of the operating unit.

O K. Ensuring that Point Beach Health Physics Procedures cnd Policies are implemented and that the WE Health Physicist is kept informed of Health Physics related activities.

6.3 TRAINING AND INSTRUCTION A health physics training program currently in operation will be applied to the steam generator repair activities. Personnel working on the i

2614Q:1 6- 4  !

steam generator project will receive instructions,and training in expo-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 activities of others who work with radioactive materials or enter radia- I tion areas will receive training. The training program includes suf-ficient instruction in the biological effects of exposure to radiation to permit the individuals receiving the instruction to understand and evaluate the significance of radiation doses in terms of the potential ri sk s. The training program also includes instruction on radiation protection rules for the plant and the applicable federal regulations.

It is planned to utilize experienced and trained personnel to implement the replacement program. The use of highly skilled craft labor should permit tasks to be performed reliably and more effeciently. Specific training sessions will be held for tasks unique to the replacement activities.

6.4 ENGINEERING AND DESIGN REVIEWS The overall steam generator replacement program will be implemented by the use of " work packages" that are amenable to efficient and timely review. These packages contain all the information required to imple-ment a specific task associated with the project, e.g. cutting the steam generator. Each of these packages is subject to intensive review, f including operation, maintenance, construction, quality assurance, health physics, and engineering personnel. The coordination in the various groups is the responsibility of the cognizant engineer. Thi s

{

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

i To the extent possible, the repair activities reflect considerations of l O personnel required to perform maintenance and inservice inspection oper-l

( ations that could lead to substantial personnel exposures. Speci- I fications for repair equipment reflect the objectives of ALARA as shown by the following examples:

l '

l t  ;

2614Q:1 6-5

A. Although the overall replaccment itself is expccted to reduce future occupational exposure, a number of specific items were also

] addressed in the specification. In addition to those identified in Section 2.0 a special device has been designed for removing primary i manway covers.

B. Special lifting lugs to facilitate handling and installation of reactor coolant pipe sections, thus minimizing exposure. l l _ ..

6.5 DESIGN FEATURES O>

'- The steam generator replacement activities involve the repair of an existing facility; therefore, the flexibility to provide features to address ALARA objectives is limited. However, to the extent possible, special provisions are being considered that will address ALARA objectives during installation and af ter the unit i s returned to service. Although the major source of personnel exposure will be from external sources, there is a potential for doses from internal expo sure. I . establishing work procedures and preparing engineering designs, the factors which detemine the doses from internal and external sources are considered. ,

For external exposures the primary concern in establishing work pro-cedures and designs is the need to limit the time personnel remain in the radiation field and the intensity of the radiation field. In order to limit the exposure time, efforts are being made to thoroughly preplan the work activity prior to its actual accomplishment, using mock-ups, n use of highly qualified individuals, use of training aids, etc. In h simple tems, the goal i s to minimize the amount of time required to complete the job. In addition to minimizing man-rem exposure, it also results in significant economic benefits.

O V

2614Q:1 6-6

The intensity of the radiation field is determined by (1) the quantity of radioactive material, (2) the nature of the emitted radiation, (3) the nature of shielding Detween the radiation source and the worker, and (4) geometry. While it is rather straight forward in limiting tne length of the stay time, it is somewhat subjective to determine how much reduction in the radiation field is cost beneficial, dach circumstance O- must be treated on a case-Dy-case basis. Methods for reducing the ra-diation fields for the steam generator replacement activities are dis-cussed in Section 3.3.5 of this report.

O Internal radiation exposure is an important consideration for the activ-ities associated with replacement because of the potential for airborne contamination attributable to cutting, welding, movement, disassembly, etc. The parameters important in determining doses from internal ex-posures are 1) the quantity of radioactive material taken into the body,

2) the nature of the material and 3) the time retained in the body.

Consequently, the basic variables that can be controlled during the repair activities to limit doses from internal exposures are those that limit 1) the amount of contamination, 2) the disposal of the con-tamination, and 3) the length of time that personnel must spend in con-taminated areas. Each situation must ce treated on a case-by-case basis. Methods for eliminating or minimizing internal doses are discussed in sections 3.3 and 3.4 of this report.

Radiation sources within a nuclear power plant differ appreciably with respect to location, intensity and characteristics. Unlike a new power plant, where the parameters affecting radiation exposure must be esti-mated using standards or experience, the radiation environment for the O steam generator replacement activities exists and can be quantified.

y/

Therefore, the environmental conditions can be determined more exactly for each location within the station. As an integral part of the plan-ning activities for the project, existing survey data, as well as new V data, are being reviewed to establish exposure levels. Once these lev-els are established, techniques can be applied to reduce them com-mensurate witn their cost-benefit.

2614Q: 1 6-7

To illustrate some of the specific measures that are being implemented and considered, each of the items in Section C.2, C.3 and C.4 of Regu-latory Guide 8.8 is addressed herein below and is numbered accordingly.

< ')

For ease of reference and comparative purposes, each of the RG 8.8 rec-ommendations is presented in tabular form in Table 6-1, with a general discussion of those measures which are tentatively planned for the steam gernirator replacement activities.

6.6 RADIOLOGICAL IMPACT 3 6.6.1 IN-PLANT DOSES The removal of the original steam generators will involve cutting along l the transition cone just below the upper girth weld. The inlet and outlet reactor coolant piping, the steam line piping, and the feedwater piping, and other miscellaneous piping will be cut to facilitate the removal of the lower shell. the upper shell of the steam generator assembly will be lifted off and placed in a storage / work location in th'e contai nraent. The lower assembly will then be lif ted from its supports and transported out of the contairraent through the equipment hatch. The replacement steam generator will then be installed in the same manner,

! only with the procedures reversed. The upper and lower assemblies will be welded together in the field. A more detailed description of these procedures is provided in Section 3.0 of thi s report.

The perfomance of the above tasks will result in doses to individuals.

Table 6-2 presents a breakdown of the estimated man-rem dose from direct l radiation exposure for the removal of the old steam generator and the

, V installation of the new steam generators. These values are best esti-mates at the present time and mcy be updated as additional survey data becomes available. The total exposure predicted for the repair is about p 1390 man-rem.

i NY

,U O

l l

2614Q: 1 6-8

6.6.2

~ DOSES TO THE PUBLIC O Due to the nature of the cutting and welding which will take place dur-ing the steam generator replacement there is the possibility of airborne particulates being generated. Steps will be taken to minimize or pre-vent this occurence, if necessary (i.e. glove boxes, tents, etc.) How-ever, since the ventilation flow in the containment will be maintained so that there will be an inward flow of air through any openings, with the exhaust through the containment purge filters, the possibility of

< releases of radioactive particulates is expected to be very small. It is also expected that there may be contaminated liquids generated during these operations associated with local decontamination, laundry, etc.

These liquids will be monitored for radioactivity; any releases will be controlled by treatment prior to discharge. Total off-site radiological dose due to replacement activities for each unit is expected to be less than that which would result if the unit were operating.

Following replacement of the steam generators and resumption of opera-tion, it is expected that there will be an over-all reduction in unit operational radiological impact due to improved steam generator performance.

l l

l O

I i

O l

O 2614Q:1 6- 9 L . _- _ _ _. _-

O O O O O O O t l TABLE 6-1

! ALARA PROVISIONS 1 of 42 I

R.G. 8.8 COMENT STEAM GENERATOR REPLACEENT PROGRAM PROVISIONS

2. Facility and Equipment Design Features

, a. Access Control of Radiation Areas a. Potential dose rates will be estimated

~

l using actual survey data, both historical

To avoid unnecessary and. inadvertent ex- and new. At the beginning of the outage i posures of personnel to radiation, the a conplete detailed survey will be taken

! magnitude of the potential dose rates at in accordance with a preplanned standard j all locations within the station should format. This survey will be periodically be measured periodically during operation updated.

to detennine current exposure potentials.

Zones associated with the higher dose rates Zones will be established in the contain-

, should be kept as small as reasonably ment work areas identifying the exposure L. achievable consistent with accessibility level in each work zone. This data will be i for accomplishing the services that must be periodically updated. Zones with high dose I perf ormed in those zones, including equip- rates will be kept as small as possible I

ment laydown requirements. Radiation zones considering the work requirements.

where station personnel spend substantial

! time should be designed to the lowest practical A detailed laydown study is now.in pro-dose rates, gress and a laydown map will be issued for the actual work.

Specific low radiation zones will be established in the containment work area which will allow personnel to take " rest breaks" without-leaving the general work area. This " rest area" will have a back-

! ground radiation exposure of 5 mR per hour or less.

A system should be established to permit effective control over personnel access to the radiation areas and control over the movement of sources of radiation within

l O O O O O O O TABLE 6-1 (Continued) 2 of 42 R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS the station. Where high radiation areas (100 mrem /hr) exist,10 CFR Part 20, 20.203 requires that station design features ,

and administrative controls provide effec-tive ingress control, ease of egress, and appropriate warning devices and notices.

Access control of radiation areas also should reflect the following considerations:

1

(1) Extraordinary design features are warranted 1. An access control point will be established

! to avoid any potential dose to personnel for the containment area at the equipment ,

that is large enough to cause acute biological hatch. Access to the containment by effects and that could be received in a short steam generator replacement personnel will period of time. Positive control of ingress be through this point. Access through the to such areas, permanent shielding, source the existing personnel hatch will be removal, or combinations of these alternatives minimized. Access through the personnel can reduce the dose potential. hatch will be controlled at the current control point location.

Part 20 requirements will be satisfied by appropriate devices, barriers, etc. Prior to the commencement of work, high radiation areas will be identified. The use of shielding and removal of source material will be implemented. As a goal, high radiation areas will be reduced to a field of less than 50 mR/hr. Where not practi-cable to do so, appropriate measures will be taken to control access to the area.

(2) Administrative controls such as standard 2. Current plant procedures will be operating procedures can be effective in pre- augmented by special procedures covering venting inadvertent exposures of personnel and steam generator replacement activities.

the spread of contamination when radioactive Appropriate " frisking" stations will De 2614Q:1

O O O O O O O TABLE 6-1 (Continued) 3 of 42 R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS material or conta.ninated equipment must be used to minimize the potential spread of transported from one station location to contamination, another and when the route of transport through lower radiation zones or " clean" areas cannot be avoided.

(3) Station features such as platforms or walk- 3. As part of the engineering and planning ways, stairs, or ladders that permit prompt ef f ort, scaff olding, hoisting and trans-accessibility for servicing or inspection of portation requirements are being identi-components located in higher radiation zones fled. Provisions will be made for the can reduce exposure of personnel who must installation of temporary scaffolding, perform these services. hoisting equipment and transportation f equipment to f acilitate steam generator removal activities.

b. Radiation Shields and Geometry Radiation shields should be designed using conservative assumptions for radioactive sourte quantities and geometries. Calculational methods known to provide reliable and accurate results (i.e., :aethods and modeling techniques that have been demonstsrated to give accept-able accuracy in analyses similar to the prob-lem of concern) should be used to detennine appropriate shield thicknesses. Shield design features should reflect the following con-siderations to maintain occupational radiation exposures ALARA:

(1) Exposure of personnel servicing a specific 1. Radiation shielding material will be used component (such as a pump, filter, or valve) where possible to minimize radiation ex-to radiation f rom other components contain- posure. This temporary shielding may con-ing radioactive material can be reduced by sist of standard components such as lead providing shielding between the individual blankets, etc., or, in some cases special 2614Q:1 -- - . - - _ - . - - . . - - _ = - - _ . .

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS components that constitute substantial shields will be fabricated. For example, radiation sources and the receptor. special shielded plugs will be installed in all openings in the steam generator lower shell assemblies prior to movement out of the containment.

(2) Where it is impracticable to provide perma- 2. The work described herein involves the nent shielding for individual components repair of an existing facility, therefore, that constitute substantial radiation sources, there is limited opportunity to change the 1 the exposure of personnel maintaining such basic plant design or layout. Temporary components can be reduced (a) by providing shielding will be provided while working as much distance as practicaDie between the on components. For example, the exposed serviceable components and the substantial piping in the reactor coolant system fol-radiation sources in the area and (b) by low cutting will be fitted with plugs providing temporary shields around components and other shielding material.

that contribute substantially to the dose rate.

(3) Potential exposure of station personnel to 3. Not applicable to repair work.

radiation from certain systems containing radiation sources can be reduced by means of a station layout that pennits the use of dis-tance and shielding between the sources and work locations. These systems include (but are not limited to) the NSSS and the reactor water cleanup, offgas treatment, solid waste treatment, and storage systems, as well as systems infrequently containing radiation sources, such as the standby gas treatment and residual heat removal systems.

(4) Streaming or scattering of radiation from 4. Streaming of radiation will be minimized locally shielded components (such as by installing shielding, such as plugs in cubicles) can be reduced by providing laby- open ended pipe lines following cutting.

't 2614Q: 1

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT FROGRAM PROVISIONS 4

rinths for access. However, such labyrinths The steam generator primary side manways or other design features of the cubicle will remain in place during the repair pro- '

should permit the components to be removed cess to eliminate streaming.

readily from the cubicle for repair or re-placement where such is work expected or antici-pated. Single-scatter labyrinths may be inadequate if the cubicle contains a sub-stantial radiation source.

(S) 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 radiation 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 inspected or serviced periodically can reduce be reused. New reflective type insula-the exposure of station personnel performing tion will be designed to provide quick these activities, and easy access to areas subject to in-service inspection.

2614Q: 1

O O O O O6 of 42 O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (8) Space within cubicles and at her shielding to 8. Not specifically applicable to repair provide laydown space for special tools and program.

ease of servicing activitiet. can reduce potential doses by permitting the services to be accomplished expeditsously, thus re-ducing exposure time.

(9) The exposure of personnel who service com- 9. While this provision is intended to apply ponents that constitute substantial radiation to permanently installed equipment, the sources or are located in high radiation general philosophy will be followed in fields can be minimized by removing the com- the repair activities. For example, the ponents and transporting them to low racia- steam generator upper shell will be l refurbished in a lower radiation area that ation zones where shielding and special tools are available. Design features that permit will be equiped with special Jib cranes the prompt removal and installation of these to f acilitate the change out of the components can reduce the exposure time. moisture separation equipment.

(10) Floor and equipment drains, piping, and 10. Not specifically applicable to repair sumps that are provided to collect and route program.

any conta.ninated liquids that might leak or be spilled from process equipment or sampling stations can become substantial radiation sourc es. The drain lines can be located in concrete floors, concrete ducts, columns, or radwaste pipe chases to prcvide shielding.

These systems can also become a source of airborne contamination because of the poten-tial f or gases to f orm in, and be re' eased by, such systems.

2614Q:1

a -

- a _ _ -:

O O O O O O 7 of 42 O .

TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS

c. Process Instrumentation and Controls Appropriate station layout and design features should be provided to reduce the potential doses to personnel who must operate service, or inspect station instrumentation and controls. The following considerations should be reflected in selecting the sta-tion features. ,

(1) The exposure of personnel who must manually 1. Not specifically applicable to the repair operate valves or controls can be reduced program.

through the use of " reach rods" or remotely operated valves or controls. However, these devices can require lubrication and maintenance that can be the source of additional exposures, and these factors ,

should be taken into consideration.

(2) The exposure of personnel wno must view or 2. While this provision is intended to apply operate instrumentation, monitors, and con- to permanently installed equipment, the trols can be reduced by locating the read- welding of the steam generator will outs or control points in low radiation zones. ,

utilize a a remote. control center for moni-toring weld parameters.

(3) Instrumentation must satisfy functional re- 3. Instrumentation will be evaluated and quirements, but the exposure of personnel appropriately selected for the specific can be reduced if the instruments are de- function to be perfonned, signed, selected, specified, and located with consideration for long service-life, ease and low frequency of maintenance and calibration, and low crud accumulation.

Operating experience should be recorded, 2614Q: 1 1 - - . , _

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS evaluated, and reflected in the selection of replacement instrumentation.

(4) The use of instrumentation that contains 4. Not specifically applicable to the repair minimal quantities of contaminated working program.

fluid, (for example, pressure transducers rather than bellows-type pressure gauges) can reduce the potential for exposure at the readout locations.

d. Control of Airborne Contaminants and Gaseous Radiation Sources

, Station design features should be provided in During the steam generator replacement all station work areas to limit the average activities, the potential for airborne concentrations of radioactive material in air contamination is increased because of the to levels well below the values listed in work tasks required in the removal process, Appendix B, Table 1, Column 1 of 10 CFR Part e.g. cutting into radioactively contaminated

20. Effective design features can minimize piping systems. Special measures will be the occurrence of occasional increases in implemented to minimize and control such air contamination and the concentrations and airborne and gaseous contamination, such as amounts of contaminants associated with any the use of temporary enclosures and filter-such occasional increases. Designs that permit ing systems. Respiratory protection equip-repeated, identified releases of large ment will be used s required to further amounts of radioactive materials into the air assure personnel saf ety.

spaces occtpied by personnel are contrary to an ALARA program.

Station design features should provide for protection against airborne radioactive material by means of engineering controls such as process, containment, and ventila-tion equipment. The routine provision of 2614Q:1 . _ . _ . - . _ _ __ . - - . . -_ , ._- - - - - ,_.

O O O .

O O O O 9 f 42 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS respiratory protection by use of individu-ally worn respirators rather than engineered design features is generally unacceptable.

The use of respirators however, might be appropriate in certain nonroutine or emer-gency operations when the application of engineering controls is not f easible or while such controls are being installed.

, The approved use of respirators is subject to Respirators will satisfy the requirements l the requirements of 10 CFR Part 20, 20.103, of 10 CFR 20, as well as R.G. 8.15 and

" Exposure of Individuals to Concentration of NUREG-0041.

Radioactive Materials in Air in Restricted Areas", and tc regulatory guidance on accept-able use. (See Regulatory Guide 8.15,

" Acceptable Programs for Respiratory Pro-tection", and NUREG-0041, " Manual of Res-piratory Protection Against Airborne Radio-active Materials". Design features of the station ventilation system and gaseous radwaste processing systems should reflect the f ollowing considerations.

I (1) The spread of airborne contamination within 1. During the steam generator replacement

! the station can be limited by maintaining activities when the equipment hatch is air pressure gradients and air flows from open, the containment ventilation system areas of low potential airborne contamination will be operated to assure that air flow to areas of higher potential contamination. is from the outside into the containment, Periodic checks would ensure that the design i.e. there will be no outleakage f rom the pressure differentials are being maintained. containment.

__ 2614Q:1__ _ _ _ _ . - _ _ __ _ _ _ _ _ . _ _ _ - . _ _ _ , _ _ _ _ - . _ . . -_ _ _ _ ___

O O O O O O O U

TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS 4

(2) Effectively designed ventilation systems and 2. Not specifically applicable to the repair gaseous radwaste treatment systems will program, contain radioactive material that has been deposited, collected, stored, or transported within or by the systems. Exposures of station personnel to radiation and to con-tamination from ventilation or gaseous rad-waste treatment components occur as a result i of the need to service, test, inspect, decon-taminate, and replace components of the sys-tems or perform other duties near these systems. Potential doses from these systems can be minimized by providing ready access to the systems, by providing space to permit the activities to be accomplished expeditiously, by separating filter banks and components to reduce exposurs to radiation from adjacent banks and components, and by providing suf-ficient space to accommodate auxiliary e ventilation or shielding of components.

(3) Auxiliary ventilation systems that augment 3. Auxiliary ventilation systems will be the permanent system can provide local con- utilized during the steam generator re-trol of airborne contaminants when equip- placement activities to provide local ment containing potential airborne sources control of airborne contaminants when is opened to the atmosphere. Two types of equipment containing potential airborne auxiliary ventilation systems have proven sources is open to the atmosphere. When to be effective. In areas where contaminated contaminated piping systems are cut they l 2614Q:1

O O O O O O 11 of 42 O

TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS equipment must be opened frequently, dampers will be surrounded by appropriate " tents" and fittings can be provided in ventilation or gloveboxes with portable blower and ducts to permit the attachment of flexible filtration equipment. A temporary con-tubing or " elephant trunks" without imbalanc- tainment ventilation system may be used to ing the ventilation system.. In areas where maintain the containment at a slightly

- contaminated equipment must be opened infre- negative pressure to assure no degradation quenetly, portable auxiliary ventilation systems of the installed system.

featuring blowers, HEPA filters, and activated charcoal filters (where radiciodine might be anticipated) on carts can be used effectively.

Portable auxiliary ventilation systems should be tested frequently to verify the efficiency of the filter elements in their mountings.

When tne efficiency has been verified, the sys-tem may be exhausted to the room or the venti-lation exhaust duct without further treatment and thus imbalance of the pennanent ventilation system can be avoided.

(4) Machining of contaminated surfaces (e.g., 4. Machining operations will be controlled welding, grinding, sanding, or scaling) or to assure that potential airborne contami-

" plugging" of leaking steam generator or nation is contained.

condenser tubes can be' substantial sources of airDorne contamination. These sources l can be controlled by using auxiliary venti-lation systems.

1 2614Q: 1

O O O O O O O 12 of 42 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (S) Sampling stations for primary coolant or 5. Not specifically applicable to the repair other fluids containing high levels of p rog ram.

radioactive material can constitute sub-stantial sources of airborne contamination.

Such sources can be controlled by using auxiliary ventilation systems.

(6) Wet transf er or storage of potentially con- 6. Not specifically applicable to the repair taminated components will minimize air program.

contamination. This can be accomplished by keeping contaminated surf aces wet, by spraying, or, pref erably, by keepirg such surf aces under water.

e. Crud control Design features of the primary coolant sys-tem, the selection of construction materials that will be in contact with the primary coolant, and features of equipment that treat primary coolant should reflect consi-derations that will reduce the production and accumulation of crud in stations where it can cause high exposure levels. The following item should be considered in the crud control effort.

(1) Production of Co-58 and Co-60, which con- 1. The materials of construction of the reactor stitute substantial radiation sources in crud, coolant system will essentially remain the can be reduced by specifying, to the extent same. Stainless steel support plates will practicable, low nickel and low cobalt bear- be used instead of carbon steel.

ing materils for primary coolant pipe, i

2614Q:1

O O O O O O O 13 of 42 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS tubing, vessel internal surf aces, heat exchangers, wear materials, and other com-ponents that are in contact with primary coolant. Alternative materials for hard f acings of wear materials of high-cobalt content should be considered where it is shown that these high-cobalt materials contri-bute to the overall exposure levels. Such consideration should also take into account potential increased service / repair require-ments and overall reliability of the new material in relation to the old. Al terna-tive materials f or high-nickel alloy materials (e.g., Inconel 600) should be considered where it is shown that these mateials contri-bute to overall exposure levels. Such con-sideration should also take into account poten-tial increased service / repair requirements and overall reliability of the new materials in relation to the old.

(2) Loss of material by erosion of load-bearing 2. Not specifically applicable to the repair hard f acings can be reduced by using program.

f avorable geometrics and lubricants, where practicable, and by using controlled leak-age purge across journal sleeves to avoid entry of particles into the primary coolant.

(3) Loss of material by corrosion can be reduced 3. The AVT chemistry program will be used in by continuously monitoring and adjusting the secondary side of the steam generators oxygen concentration and pH in primary and technical specifications maintained for coolant above 250*F and.by using bright the primry coolant chemistry.

hydrogen-annealed tubing and piping in the primary coolant and feedwater systems.

2614Q:1

O O O O O O O 14 of 42 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REDLACEMENT PROGRAM PROVISIONS

, (4) Consideration should be given to cleanup 4. No changes to the reactor water cleanup systems (e.g., using graphite or magnetic systems are planned. Based on the results filters) f or removal of crud f rom the primary of studies conducted for the repair pro-coolant during operation' gram, it is not planned to perform any chemical decontamination. Also, as a

result of these studies, flushing of the systems will provide no significant bene-1 fit in reducing radioactive exposure.

(5) Deposition of crud within the primary 5. The new lower assemblies will have flush coolant system can be reduced by providing tubes welded in the tube sheet. The tube laminar flow and smooth surf aces for coolant ends presently project below the bottom of and by minimizing crud traps in the system the tube sheet. This change effectively to the extent practicable. reduces entrance losses and eliminates a potential crud trap. The replacement lower assemblies will be equipped with an improved blowdown system to mirlisiize crud deposition. The overall design of the replacement steam generators has as a design objective to minimize crud deposition. See Section 2.0 of this report for more details.

f. Isolation and Decontamination Potential doses to station personnel who must service equipment containing radioactive sources can be reduced by removing such sources f rom the equipmenet (decontamination) to the extent practicable, prior to servic-i ng . Serviceable systems ano components 2614Q:1

[ o o o o o o o TABLE 6-1 (Continued) 1

! R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS that constitute a substantial radiation

) source should be designed, to the extent j practicable, witn features that permit iso-a lation and decontamination. Station design i features should consider, to the extent practicable, the ultimate decommissioning 4

of the f acility and the f ollowing concerns:

(1) The necessity for decontamination can be 1. Drains will be provided in the steam reduced by limiting, to the extent practica- generator channel heads to allow all ble, the deposition of radioactive material water to be removed before entrance for within the processing equipment--particularly maintenance and/or inspection.

in the " dead spaces" or " traps" in components j where substantial accumulations can occur.

The deposition of radioactive material in piping can be reduced and decontamination efforts enhanced by avoiding stagnant legs, by locating connections above the pipe center-line, by using sloping rather than horizontal runs, and by providing drains at low points in the system.

(2) The need to decontaminate equipment and 2. The replacement of the steam generators

station areas can be reduced by taking will reduce the probability of radio-i measures that will reduce the probability active releases, as well as the amount of release, reduce the amount released, and released.

i reduce the spread of the containment f rom the source (e.g., from systems or com-ponents that must be opened for service or replacement) . Such measures can include auxiliary ventilation systems (see 4.b),

2&D42:1

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS treatment of the exhaust f rom vents and j overflows (see 2.h(8)), drainage control such as curbing and floors sloping to l local drains, or sumps to limit the spread of contamination f rom leakage of liquid systems.

(3) Accumulations of crud or other radioactive 3. Not specifically applicable to the repair material that cannot be avoided within com- p rog ram.

ponents or systems can be reduced by pro-viding features that will permit the recir-culation or flushing of fluids with the capacity to remove the radioactive material through chemical or physical action. The fluids containing the contaminants will require treatment and this source should be considered in sizing station radwaste treat-ment systems.

(4) Continuity in the f unctioning of processing 4. Not specifically applicable to the repair ]

or ventilation systems that are important p rog ram.

for controlling potential doses to station i personnel can be provided during servicing of the systems if redundant. components or systems are available so that the component (with associated piping) being serviced can be isolated. I (5) The potential for contamination of " clean 5. Not specifically applicable to the repair services" (such as station service air, p rog ram.

nitrogen, or water supply) f rom leakage f rom 2614Q:1 ..

TABLE 6-1,(Continued) 17 of 42 R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS adjacent systems containing contaminants l can be reduced by separating piping for

these services f rom piping that contains

! radioactive sources. Piping that carries radioactive sources can be designed for the

lifetime of the station, thus avoiding the necessity for replacement (and attendant exposures) and lessening the potential f or contamination of clear services if it is impracticable to provide insolation through separate chases.

. (6) Surf aces can be decontaminated more expe- 6. Not specifically applicable to the repair i ditiously if they are smooth, nonporous, program.

and f ree of cracks, crevices, and sharp corners. . These desirable features can j.. be realized by specifying appropriate design instructions, by giving attention i to finishing work during construction or

, manuf acture, and by using sealers (such -

l as special paints) on surf aces where 1

contamination can be anticipated. (ANSI N-101.2 provides helpf ul guidance on this matter).

(7) Where successf ul decontamination of impor- 7. Not specifically applicable to the repair tant systems could be prevented by an anti- program.

cipated f ailure of a critical component or feature, additional features that permit i

alternative decontamination actions can be i

provided.

1 i

2614Q:1

O O O O O O O TABLE 6-1 (Continued) 18 of 42 R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (8) Contaminated water and deposited residues 8. Not specifically applicable to the repair in spent f uel storage pools contribute prog ram, to the exposure at accessible locations in the area. Treatment systems that remove contaminants f rom the water can perf orm more efficiently (a) if intake and dischage poir.i.s for the treatment systems are locat ad to provide enhanced mixing and to avoid stagnation areas in the pool and (b) if pool water overflows and skimmer tanks are provided. Fluid jet or vacuum cleaner type agitators car, help reduce the settling of crud on surf aces of the pool system.

g. Radiation Monitoring Systems g. Not specifically applicable to the repair program.

Central or " built-in" monitoring systeras that give inf ormation on the dose rate and concentration of airborne radioactive material in selected station areas can i reduce the exposure of station personnel who would be required to enter the areas to obtain the data if such systems were not provided. These systems also can provide timely information regarding changes in the dose rate or concentrations of airborne radioactive material in the areas. (The installation of a central monitoring system is easier and less

! expensive if it is a part of the original l station design.) The selection or design i and installation of a central monitoring 2614Q:1

O O O O '

O O O 19 of 42 TABLE 6-1 (Continued)

I R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS system should include consideration of the following desirable features:

(1) Readout capability at the main radiation protection access control point; (2) Placement of detectors for optimum coverage of areas;

(3) Circuitry that indicates component f ailure; l (4) Local alarm and readout;

(5) Clear and unambiguous readout; (6) Ranges adequate to ensure readout of the highest anticipated radiation levels and to ensure positive readout at the lowest anticipated levels; and (7) Capability to record the readout of all j systems.

i l h. Resin and Sludge Treatment Systems h. Not specifically applicable to the repair program.

Systems used to transport, store, or process resins or slurries of filter sludge present a special hazard because of the concentrated nature of the radioactive material. Design features for resin and sludge-handling systems should reflect this concern and the I following specific considerations:

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (1) The accumulation of radioactive material 1. Not specifically applicable to the repair in components of systems used to process p rog ram.

l resin and sludges can be reduced by (a) Heducing the length of piping runs; (b) Using larger diameter piping (to minimize plugging);

(c) Reducing the number of pipe fittings; (d) Avoiding low points and dead legs in piping ;

(e) Using gravitational flow to the extent practicable; and (f ) Minimizing flow restrictions of processed material.

(2) The need for maintenance and the presence 2. Not specifically applicable to the repair of intense local radiation sources can be reduced by:

(a) Using f ull-ported valves constructed such that the slurry will not interfere with the opening or closing of the valve and (b) Avoiding cavities in valves.

(3) The deposition of resin and sludge that 3. Not specifically applicable to the repair would occur if elbow fittings were used program.

can be reduced by using pipe bends of at least five pipe diameters in radius.

2614Q: 1

O O O O O 21 of 42 O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS' i l Where pipe bends cannot be used, long radius elbows are preferred.

(4) Smoother interior pipe surf aces at con- 4. Not specifically applicable to the repair

! nections (with attendent reductions in program.

f riction losses, deposition of material, and tendencies to " plug") can be achieved i by using butt welds rather than socket welds and by using consumable inserts 4 rather than backing rings. "

(5) Where the use of tees cannot be avoided, 5. Not specifically applicable to the repair i line losses can be reduced if the flow prog ram.

is through the run (straight section) of the tee, and accumulations of material i in the branch of the tee can be reduced by orienting the branch horizontally or (pref erably) above the run.

(6) Slurry piping is subject to plugging that 6. Not specifically applicable to the repair may require backflushing from the tank and program.

equipment isolation valves and pressurizing with water, nitrogen, or air to " blow out"

pl ugged lines. However, the use of pres-surized gas for blowing out lines can pre-sent a potential contamination source and may not be effective in relieving plugged lines.

l (7) Water, air, or nitrogen for spargirg can 7. Not specifically applicable to the repair I be used to fluidize resins or sludges program.

( in storage tanks. The use of gases, however, presents a potential source of airborne contamination and tank rupture f rom overpressures.

l l

4 2614Q:1

O O O O O O O TABLE 6-1 (Continued) 22 of 42 R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (8) The spread of contamination by the loss of 8. Not specifically applicable to the repair resin or sludge through overflows and program.

vents can be reduced by using screens, filters, or other features that will collect and retain solids. However, such features generally require cleaning by remote flushing, by rapid replacement, or by other means to reduce exposures during servicing .

I 1. Other Features Station layout and station tasks should be reviewed to identify and provide special features that complement the ALARA Program.

Station design should reflect consideration of the following concerns:

(1) The selection of radiation-damage-resistant . 1. Materials used satisfy this criteria.

materials for use in high radiation areas can reduce the need f or f requent replacement and can reduce the probability of contamination f rom leakage. ,

(2) The use of stainless steel for constructing 2. Not specifically applicable to the repair

)

or lining components, where is is compatible prog ram.

with the process, can reduce corrosion and can provide options for decontamination methods.

2614Q:1

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (3) Field-run piping that carries radioective 3. Detailed piping layout will be provided for material can cause unnecessary exposures the steam generator replacement activities.

unless due consideration is given to the routi ng . Such unnecessary exposures cc.n be avoided if the routing is accomplished under the cognizance of an individual f amiliar with the principles of radiation protection or if a detailed piping layout is provided, i.e., if the piping is not f iel d-run.

(4) Where filters or other serviceable components 4. Not specifically applicable to the repair can constitute substantial radiation sources, prog ram, exposures can be reduced by providing features that permit operators to avoid the direct radiation beam and that provide remote removal, installation, or servicing.

Standardization of filters should be considered.

(5) The servicing of valves can be a substantial 5. Not specifically applicable to the repair source of doses to station personnel. These prog ram.

doses can be reduced by providitg adequate working space for easy accessibility and by locating the valves in areas that are not in high radiation fields.

(6) Leakage of contaminated coolant f rom the 6. Not specifically applicable to the repair primary system can be reduced by using live- p rog ram.

1oaded valye packings and bel 1ow seals.

unna3

O O O O O O O

, TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS i

(7) Potential doses from servicing valves and 7. Not specifically applicable to the repair i

from leakage can be reduced by specifying prog ram.

and installing reliable valves for the required service, by using radiation-damage-resistant seals and gaskets, and i by using valve back seats. The use of

! straight-through valve configurations can avoid the buildup of accumulations in internal crevices and the discontinuities that exist in valves of other configurations.

In most cases, valves can be installed in the

" stem-up" orientation to f acilitate mainte-nance and to minimize crud traps. The desired f eatures are reliability, good per-formance, and the ability to be maintained infrequently and rapidly.

(8) Leaks f rom pumps can be reduced by using 8. Not specifically applicable to the repair canned pumps where they are compatible program.

with the service needs, provided that lower personnel exposures can be achieved thereby.

If mechanical seals are used on a pump in a slurry service, features that permit the use of flush water to clean pump seals can reduce the accumulation of radioactive material in the seals. Drains on pump housings can reduce the radiation field from this source during servicing. Provision for the collection of such leakage or disposal to a drain sump is appropriate.

(9) The sources of radiation such as sedimentation 9. Not specifically applicable to the repair that occurs in tanks used to process liquids prog ram.

containing radioactive material and residual liquids can be reduced when servicing by draining the tanks. The design can include 2614Q:1

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8' COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS i

slopiry the tank bottoms toward outlets leading to other reprocessing equipment and, where practicable, providing built-in spray or surge features.

(10) Spare connections on tanks or other cogonents 10. Not specifically applicable to the repair located in higher radiation zones may be program.

desirable to provide flaxibility in operations.

Exposures of personnel can be avoided if these connections are provided as a part of the i

original equipment rather than by subsequent

! modification of the equipment in the presence of radiation.

(11) Inspections to satisfy the ASME Code and 11. Appropriate provisions will be implemented regulatory requirements can result in to minimize exposure of station personnel exposures of station personnel to radiation. in performing Code inspections, such as Many of the objectives presented above will removable insulation, smooth welds, etc. i aid in reducing potential exposure: to per-sonnel who perform the required inspections.

Station features and design should, to the '

extent practicable, permit inspections to be accomplished expeditiously and with minimal exposure of personnel. The ALARA effort can also be aided by prompt accessibility, shielding and insulation that can be quickly removed and reinstalled, and special tools and instruments that reduce exposure time or permit remote inspection of components or equipment containing potential radiation sources.

2614Q:1

O O O O O O O TABLE 6-1 (Continued) 26 of 42 R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (12) Components can be removed from processing 12. Not specifically applicable to the repair systems more expeditiously if adequate program.

space is provided in the layout of the system and if the interconnections permit prompt disconnects.

(13) Station features that provide a favorable 13. Temporary lighting and scaffolding will be working environment, such as adequate installed to provide a favorable work lighting, ventilation, working space, and environment.

accessibility (via such means as working platforms, cat walks, and fixed ladders),

can promote work efficiency.

(l;[) The exposura of station personnel who must 14. Not specifically applicable to the repair

_ replace lamps in high radiation areas can program.

be reduced by using extended service lamps and by providing design features that permit the servicing of the lamps from lower radiation areas.

(15) An adequate emergency lighting system can 15. An emergency lighting system will be reduce potential exposures of station available for the steam generator personnel by permitting prompt egress from replacement activities.

high radiation areas if the station lighting system fails. ,

2 l

2614Q: 1

O O O O O O O 27 of 42 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS

3. Radiation Protection Program A substantial portion of the radiation dose This section of the ALARA guidelines is to station personnel is received while they directly applicable to the steam genera-are perf orming services such as maintenance,- tor replacement activities and the philo-ref ueling, and inspection in high radiation sophy identified will be carried out.

areas. The objectives that were previously Many efforts are being irrplemented to presented in Section 2(RG 8.8) can provide assure that the number of persons who station design features conducive to an effec- must enter high radiation areas or con-tive ALARA Program. However, an effective taminated areas are held to a minimum; ALARA Program also requires station opera- that the period of time the persons must tional considerations in terms of procedures, remain in these areas is minimized; and job planning, record keeping, special that the magnitude of the potential dose equipment, operating philosophy, and other is maintained to the lowest levels support. This section deals with the manner commensurate with other considerations.

in which the station adninistrative efforts can influence the variable of (1) the number of persons who must enter high radiation areas or contaminated areas, (2) the period of time the persons must remain in these areas, and (3) the magnitude of the potential -

dose.

a. Preparation and Planning a. Each task will be identified as a " work package" which will contain a specific pro-Bef ore entering radiation areas where cedures, drawings, instructions, shelters significant doses could be received, station and other pertinent inf ormation necessary personnel should have the benefit of prepara- to perform the task. These plans will tions and plans that can ensure that exposures be available prior to the actual com-are ALARA while the personnel are performing mencement of the work and will have the services. Preparations and plans should undergone extensive review by engineering, reflect the following considerations: operations, health physics and construc-tion personnel.

51@iB

O O O O O O 28 of 42 O

TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (1) A staff member who is a specialist in 1. A staff member qualified and experienced radiation protection can be assigned the in radiation protection has been assigned responsibility for contributing to and full time to assist in preparing the coordinating ALARA efforts in support of " work packages". During the actual operations that could result in substantial replacement activities, he will be respon-individual and collective dose levels. sible for all radiation protection activi-ties, including dose control and monitoring, issuance of radiation work permits, etc.

The specific responsibilities are identified elsewhere herein.

(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 l

infonnation with respect to radiation, con- been reviewed. Supplemental surveys will tamination, airborne radioactive material, . have also been performed. 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 thereafter. Previous operating experience has been reviewed regarding mechanical difficulties and this information is being incorporated into the work pack-ages. Since the original installation of 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.

l 2614Q: 1

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMfENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIOhS (3) Radiation surveys provided in conjunction 3. Radiation surveys will be taken at the with inspections or other activities can commencement of the outage and periodi-define the nature of the radiation fields cally thereaf ter, as well as special and identify f avorable locations where surveys for specific tasks. The work personnel may take advantage of available areas, temporary shielding, etc. are shielding, distance, geometry, and other being planned considering expected f actors that affect the magnitude of the exposure levels.

dose rate or the portions of the body

• exposed to the radiation.

(4) Photographs of "as installed" equipment or 4. Photographs have been taken for planning components can be valuable f or planning purposes. In addition, photographs have purposes and can be augmented by additional Deen utilized in the engineering and plan-photos taken during the surveys. The use ning eff ort. It is planned to maintain a of portable TV cameras with taping features photographic record of work in the pro-has considerable merit as both an opera- gram. Video taping is being considered tional aid and a teaching aid. f or recording certain tasks.

(5) The existing radiation levels f requently S. An extensive evaluation of possible decon-can be reduced by draining, flushing, or tamination techniques has been performed Other decontamination methods or by and is discussed in Section 3.4 of this i

removing and transporting the component rep ort.

to a lower radiation zone. An estimate of the potential doses to station personnel expected to result from these procedures is germane in selecting among alternative actions.

2614Q:1 . . _ . _ _ _ _ . . _ _ _ . __ _ _ _

O O O O O O O 30 of 42 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (6) A pre-operational briefing for personnel 6. Preoperation briefings will be performed who will perform pervices in a high radia- to instruct personnel on the requirements i i tion area can ensure that service personnel -

of specific work packages. As a minimum, understand the tasks about to be performed, each " work package" will be reviewed by l the information to be disseminated, and the the personnel performing the work. The l special instructions to be presented. work packages provided are self explana-J tory and provide sufficient information to perform the work.

(7) A program can be implemented to provide 7. An access control program will be in access control and to limit exposures to effect during the replacement activities.

those persons needed to perform the Specifically an access control point will required services in the radiation areas. be established at the entrance to the Such a program would address conditions work area, i.e. equipment hatch that require a special work permit or other and will be manned by a radiation pro-special procedures, tection staff memoer. Additional radia- '

tion protection personnel will be avail-able to monitor such assignments as

, required. The "Radiaion Work Permit" t

(RWP) system presently in use at the plant will be used.

(8) A work permit form with an appropriate format 8. A " Radiation Work Permit" System is pre-can be useful for recording pertinent sently being used at the plant. This

information concerning tasks to be performed system is presently being reviewed for in high radiation areas so that the informa- used on the steam generator replacement i

tion is amenable to cross-referencing and activities, primarily as a result of the statistical analysis. Information of interest desirability to make the data obtained would include the following items: amenable to statistical analysis. The 1

2614Q:1

O O O O O O 31 of 42 O

TABLE 6 _1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS

' information listed in items (a) through (a) Designation of services to be performed (j) will typically be recorded for the work on specific components, equipment or activities.

systems; (b) Number and identification of personnel working on the tasks; (c) Anticipated radiation, airborne radio-active material, and contamination levels, based on current surveys of the work areas, and date of survev; (d) Monitoring regnirements, such as con-tinuous air monitoring or sampling equipment; (e) Estimated exposure time required to com-plete the tasks an:1 the estimated doses anticipated from the exposure; (f) Special instructions and equipment to minimize the exposures of personnel to radiation and contan:ination; (g) Protective clothing and equipment re-quirements; (h) Personnel dosimetry requirements; (i) Authorization to perform the tasks; and (j) Actual exposure time, doses, and other infonnation obtained during the opera-tion.

(9) Consideration of potential accident situations 9. Where appropriate " work packages" will or unusual occurrences (such as gross conta- address potential accident situations mination leakage, pressure surges, fires, cuts or unusual occurrences and will include punctures, or wounds) and contingency planning appropriate contingency planning.

2614Q: 1

O O O O O O 32 of 42 O

TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS can reduce the potential for such occurrences and enhance the capability for coping with the situations expeditiously if they occur.

I (10) Portable or temporary shielding can reduce 10. Temporary shielding, e.g. lead blankets 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-performed. ing of the reactor coolant piping, shield-ing will be used to isolate those portions of the system whicn.are contaminated.

(11) Portable or temporary ventilation systems 11. Temporary ventilation systems will be 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-contamination and limit the intake by priate, coverings will be used to minimize workers through inhalation. spread of contamination.

(12) " Dry runs" on mock-up equipment can be 12. Dry runs and mock-up equipment will be useful for training personnel, identifying used for training personnel and testing problems that can be encountered in the equipment. For example, a full scale actual task situation, and selecting and mock-up of the channel head and the transi-qualifying special tools and procedures tion cone areas will be used to simulate to reduce potential exposures of station the welding of the generator. The actual personnel. equipment to be used at the site will be used for this demonstration. Welding technicians will also be trained on the mock-up. Special handling tools and equipment will be used extensively.

)

2614Q:1

O O O O O O O 0 4 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS For instance, a special handling and transi-port system will be installed for~ handling the steam generator lower assemblies during removal and installation. Temporary job cranes will be installed in the con-tainment to f acilitate the handling of the moisture separation equipment and other equipment. Trash compactors will be in-stalled in the containment to minimize waste volume and to reduce handling.

Automatic welding equipment will be used f or welding the reactor coolant piping and will be used to train personnel prior to the actual work. Many other similar pro-visions will be implemented to reduce exposure times.

(13) Adequate auxiliary lighting and a comfortable 13. Auxiliary lighting will be provided as environment (f or example, vortex tube coolers required. The specific requirements have f or supplied air suits) can increase the effi- not yet been established. The containment ciency of the work and thus reduce the time environment should be comfortable without spent in the higher radiation zones. any significant changes since the reactor will be def ueled thus eliminating the major heat sourt:e. Because the contain-ment is f ully enclosed by concrete walls, the ambient working conditions in winter

- and summer are expected to be comf ortable, e

Operation of the temporary containment ventilation system will assure that there is adequate air flow in the containment.

2614Q: 1

O O O O O O O 34 UI 42 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (14) Radiation monitoring instruments selected 14. Radiation monitoring equipment of the type and quantity required to monitor and made available in adequate quantities ,

can permit accurate measurements and radiation and contamination levels will rapid evaluations of the radiation and be available. Calibration of the monitor-contamination levels and changes in levels ing equipment will be in accordance with when they occur. Routine calibration of

• the Point Beach Quality Assurance Manual.

instruments with approprieate scurces and testing can ensure operability and accuracy of measurements.

(15) Performing work on some components inside 15. Temporary tents and glove boxes will be disposable tents or, for less complicated used where appropriate, e.g. cutting of jobs, inside commercially available dis- the reactor coolant piping.

posable clear plastic glove bags can limit the spread of contamination. Such measures can also avoid unnecessary doses resulting from the need to decontaminate areas to permit personnel access or to allow for entry with less restrictive protective clothing and equipment requirements.

(16) Careful scheduling of inspections and other 16. The entire outage will be scheduled in tasks in high radiation areas can reduce ex- detail, commensurate with other require-exposures by permitting decay of radiation ments, to take advantage of radioactive sources during the reactor shutdown period decay considerations. Data from and by eliminating some repetitive surveys. historical surveys and experience is being Data from surveys and experience attained .in utilized in the planning effort.

previous operations and current survey data can be factored into the scheduling of specific tasks.

2614Q: 1

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

35 of &

TABLE 6-l'(Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS ,

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 TLS's will direct-reading dosimeters, alarming dost- 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 of those doses to specific operations.

2614Q:1

t O O O O O O 36 of 42 O

TAdLE 6-1 (Continued) l R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (3) Communications systems between personnel 3. Audio communication systems will be in high radiation zones and personnel who available for use during the outage i are monitoring the operation in other loca- activities. A video communications i tions can permit timely exchanges of informa- system is being pursued.

tion and avoid unnecessary exposures to monitoring personnel. ,

c. Postoperations Observations, experience, and data obtained during nonroutine operations in high-radia-tion zones should be ascertained, recorded, and analyzed to identify deficiencies in the progran and to provide the bases for revising procedures, modifying features, or making l

other adjustments tnat may reduce exposures i during subsequent similar operations.

l (1) Formal or informal postoperation debriefings 1. Postoperational debriefings will be used

? of station personnel performing the services to obtain infonnation regarding the can provide valuable information concerning requirements of the work packages.

shortcomings in pre-operational briefings, -

planning procedures, special tools, and other factors that contributed to the cause of

}

doses received during the. operation.

(2) Dose data obtained during or subsequent to 2. The data listed in 3.a.8 (RG 8.8) will an operation can be recorded in a preselected typically be recorded and analyzed.

! manner as part of a " Radiation Work Permit" or i similar program (see 3.a(8), RG 8.8) so that

' the data are amenable to statistical analyses.

l l

I 2614Q: 1 ,

O O O O O O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (3) Information concerning the cause of compo- 3. The purpose of the replacement is to nent failures that resulted in the need for accomplish this objective.

. servicing in high radiation areas can provide a basis for revising specifications on replace-ment equipment or for other modifications that can improve the component reliability.

! Such improvements can reduce the frequency of servicing and thus reduce attendant exposures.

i (4) Information gained in operations can provide 4. To the extent practicable, information

! a basis for modifying equipment selectior. gained from the replacement operation will and design features of new facilities. be considered in future designs.

(5) Summaries of doses received by each category 5. The exposure data obtained will be com-of maintenance activity can be reviewed pared with the estimated exposures.

periodically by upper management to compare the incremental reduction of doses with the

• cost of station modifications that could be

! made.

4. Radiation Protection Facitilities, Instru-mentation, and Equipment j A radiation protection staff with facilities, Additional radiation monitoring equipment, 1 instrumentation, and protective equipment protective clothing and related items will J adequate to permit the staff to function be furnished for the steam generator efficiently is an important element in replacement activities.

9 achieving an effective ALARA program. The i

selection of instrumentation and other -

equipment and the quantities of such equip-ment provided for normal station operations should be adequate to mcat the anticipated l

, 2614Q:1

O O O .

O O O 38 of 42 O

TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS needs of the station during normal operations and during major outages that may require supplemental workers and extensive work in' high radiation areas. (Accident situations are not considered in this guide.) Station design features and provisions should reflect the following considerations.

,a. Counting Room A low-radiation background counting room is needed to perform routine analyses on station samples containing radioactive ,

material collected from air, water, surfaces, i and other sources. An adequately equipped counting room wouild include (1) Multichannel gamma pulse height analyzer 1. The existing counting room is equipped with

, (Regulatory Guide 5.9, " Specifications for a multichannel analyzer which will be Ge(Li) Spectroscopy Systems for Material utilized.

. -Protection Measurements--Part 1: Data i

Acquisition Systems", provides guidance for selecting Ge(Li) spectroscopy systems);

(2) Low-background alpha-beta radiation pro- 2.. These counters are available.

portional counter (s) or scintillation coun-ter(s);

(3) End-window Geiger-Muller (G-M) counter (s); 3. These counters are available.

and 2614Q:1

~

~

O O O O O 39 of 42O O TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (4) A liquid scintillation counter f or tritium 4. Equipment available.

analysis. Ainalyses of bioassay and environ-mental samples and whole body counting (see Regulatory Guide 8.9 " Acceptable Concepts, Models, Equations, and Assunptions for a Bioassay Program") call f or additional equipment and laboratory space if the analyses are performed by station personnel rather than by other specialists through contractual arrangements.

b. Portable Instruments b.

Portable instruments needed for measuring Appropriate equipment is available.

dose rates and radiation characteristics would include (1) Low-range (nominally 0 to 5 R per hour) ion chambers or G-M rate meters; (2) High-range (0.1 to at least 500 R per hour) ion chambers; (3) Alpha scintillation or proportional count rate meters; (4) Neutron dose equivalent rate meters; (5) Air samplers for short-term use with particulate filters and iodine collec-tion devices (such as activated charcoal cartridges); and (6) Air monitors with continuous readout features.

l - _ .

2614Q:1

! O O O O O O O

! TABLE 6-1 (Continued) l R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS -

c. Personnel Monitoring Instrumentation c.

Personnel monitoring instrumentation selections Appropriate equipment is available.

should include consideration of (1) G-M " Friskers" for detecting low level of radioactive material;

.(2) Direct-reading low-range (0 to 200 mR) and intermediate-range (0 to 1000 mR)

pocket dosimeters (see Regulatory l

Guide 8.4);

(3) Alam Dosimeters; (4) Film badges and/or thermouminescent dosimeters (TLD); t (5) Hand and foot monitors; and i

(6) Portal monitors.

1

'l

d. Protective Equipment Utility-supplied protective equipment selec-

tion should include consideration of

(1) Anti-contamination clothing and equipment 1. Anti-contamination clothing will be used.

that meet the requirements of ANSI Z-88.2, 1969 for use in atmospheres containing radioactive materials, or the National i Institue of Occupational Safety and Health's (NIOSH) " Certified Personal

, Protective Equipment List", July 1974, and current supplements fro DHEW/PHS.

i 2614Q:1

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

41 of 4 TABLE 6-1 (Continued)

R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS (2) Respiratory protective equipment including 2. Respiratory protective equipment is a respirator fitting program that satisfies available. Also, a respirator fitting the guidance of Regulatory Guide 8.15 and program that satisfies the guidance of NUREG-0041. Regulatory Guide 8.15 and NUREG-0041 is available and snall be utilized as acceptable.

e. Support Facilities Design features of r'adiation protection support facilities should include consideration of:

! (1) A portable-instrument calibration area 1. Calibration area suitable for calibration designed and locted such that radiation in is used.

the calibration area will not interfere with low level monitoring or counting systems; (2) Personnel decontamination area (this f acility 2. A special decontamination area is being should be located and. designed to expedite provided for the use of personnel assigned ,

rapid cleanup of personnel and should not be to the steam generator replacement project, used as a multiple-purpose area or share and will be equipped with the appropriate ventilation with food-handling areas) with facilities. The existing plant facili-showers, basins, and installed "frister" ties will also be used if required.

equipment; (3) Facilities and equipment to clean, repair, 3. The existing plant decontamination

' and decnntaminate personnel protectiave equip- facility will be used to clean, repair ment, monitoring instruments, hand tools, and decontaminate personnel, protective electromechanical parts, or other material clothing, hand tools, etc. The need to (highly contaminated tools or other equipment augment cleaning of protective clothing should not be decontaminated in the area is being pursued.

used to clean respiratory equipment) 2614Q: 1

42 of 4 TABLE 6-1 (Continued) l R.G. 8.8 COMMENT STEAM GENERATOR REPLACEMENT PROGRAM PROVISIONS l

l 4. As noted in ited 2, a special personnel j (4) Change rooms that (preferably) connect with the personnel decontamination area and a change facility will ne provided for control station area equipped with sufficient steam genertor replacement unit and will i lockers to accommodate permanent and contract be equipped with the approprieate facili-maintenance workers who may be required ties such as lockers.

i j during major outages; 4

(5) Control stations for entrance or exit of 5. A control station will be established at l personnel into radiation and contamination the containment equipment hatch. Ingress l

controlled access areas of the station, such and egress for the containment work area as the personnel entrance 1.0 the containment will be through this station. The personnel buildings and the main entrance to the hatch will be used minimally as necessary.

radwaste processing areas; these control l

stations also may be used as the control i point for radioactive material movements l

throughout the station and for the

- storage of portable radiation survey equip-ment, signs, ropes, and respiratory protec-tive equipment;

6. Audio communications systems will be

) (6) Equipment-to facilitate communication between all areas throughout the station; available for use.

l i

(7) Sufficient office space to accommodate the 7. The area identified in items 2 and I

I temporary and perranent radiation protection 4, as well as a temporary construction, will contain office space to

! staf f, permanent records, and technical accomodate the steam generator project l literature.

staff.

i 2614Q: 1

f Page 1 of 8 l

O TABLE 6-2 ,

t ESTIMATE OF PERSONNEL RADIATION EXPOSURES FOR }

O STEAM GENERATOR REPLACEMENT OPERATIONS

{

AT P0 INT BEACH UNIT - I l PHASE-I SHUTDOWN AND PREPARATORY ACTIVITIES I

ESTIMATED ESTIMATED  !

TASK LABOR EXPOSURE l DESCRIPTION (MAN-HOURS) (MAN-REM)  ;

I Shutdown and Preparatory 58,887 237.3 }

Activities i

II Removal Activities 141,680 421.7 l III Installation Activities 334,138 605.8 IV Post Installation and 87,700 118.3 l Startup Activities l V Steam Generator Storage 1,532 6.6 f Activities PROJECT TOTALS 623,937 1389.7

, ( All Tasks)  :

(

O '

!O  !

O .

i 2614Q: 1 f

Page 2 of 8 O

i TABLE 6-2 ESTIMTE OF PERSONNEL RADIATION EXPOSURES FOR

^

i STEAM GENERATOR REPLACEMENT OPERATIONS AT POINT BEACH UNIT - I -

l PHASE-I SHUTDOWN AND PREPARATOR( ACTIVITIES  ;

O -,

l; ESTIMATED ESTIMATt0 TASK LABOR EXPOSURE DESCRI PTION (MN-riOURS) (MN-REM)  !

t Install Polar Crane Gibpole 2,000 10.0 Modification .

Installation of Jib Cranes 5,584 6.7 l Misc. Disassemble Manipulator 1,000 2.0 f Crane ,

i Install Steam Generator 5,738 6.7 l Transport System  !

Removal Constainment 2,000 3.5 ,

Obstructions ,

I Installation of Temporary 2,000 2.5  !

Ventilation System Temporary Scaffolding 5,000 29.7 Temporary Lighting and Power 2,000 2.0 Cleanup and Decon 10,712 35.0 -

Polar Crane Operator 1,000 2.0 Shielding 11,500 100.0 O  :

2614Q:1 i

i

Page 3 of 8 O TABLE 6-2 ESTIMTE OF PERSONNEL RADIATION EXPOSURES FOR STEAM GENERATOR REPLACEENT OPERATIONS AT POINT BEACH UNIT-I PHASE-1: SHUTDOW AND PREPARAT0RI ACTIVITIES O

ESTI E TED ESTI E TED TASK LABOR EXPOSURE DESCRI PTION (MN-HOURS) (EN-REM)

H.P., Q.A 12,723 15.0 Miscellaneous 2,000 5.0 Installation of Service 630 2.2 O Ai r System 2,000 Work Platform 1.0 Modification Protection of Containment 1,500 8.0 Components Project Supervision and 1,000 0.0

Admini stration SUBTOTAL PHASE I 58,d87 237.3 O

O O

2614Q:1

~~.

Page 4 of 8 TABLE 6-2 ESTIl%TE OF PERSONNEL RADIATION EXPOSURES FOR STEAM GENERATOR REPLACEMENT OPERATIONS l AT POINT BEACH UNIT-1 PHASE II: REMOVAL ACTIVITIES '

ESTIMATED ESTIMATED TASK LABOR EXPOSURE ,

DESCRI PTION (MAN-HOURS) (MAN-REM) l Removal of Insulation 1,224 9.3 L (lower shell, RC piping)  !

Removal of Insulation 446 3.2 (upper shell, mainstream  :

I and feedwater piping)

Removal of Miscellaneous 3,356 24.4 Piping Set Up Steam Generator 600 2.0 i Girth Cut Equipment Cut and Remove Steam 3,536 6.8 Generator Upper Shell Cutting of Reactor 9,139 96.9 Coolant Piping Cutting of Mainstream and 1,412 2.4 Feedwater Piping Di sassembly of Steam 5,910 34.7 f Generator Supports Removal of Moisture 3,794 8.1 Separation Equipment Refurbish Steam 11,543 10.2  !

Generator Upper Shell  ;

Removal of Steam 1,892 4.7 Generator Level Instru-ments and Blowdown Piping O Removal of Steam Generator Lower Shell 2,733 17.6  !

l l

l 2614Q:1

l Page 5 of 8 l O TABLE 6-2  ;

ESTIMATE OF PERSONNEL RADI ATION  !

EXPOSURES FOR STEAM GENERATOR REPLACEMENT OPERATIONS AT  !

POINT BEACH UNIT-1 l PdASE II: REMOVAL ACTIVITIES l O  :

l

?

I ESTIMATED ESTIMATED f TA3K LABOR EXPOSURE  !

DESCRI PTION (MAN-HOURS) (MAN-REM) i Temporary Scaffolding 8,227 29.1 Temporary Lighting and Power 3,810 5.2 [

Cleanup and Decon 35,731 86.2 r

Polar Crane Operator 1,837 1.7 l H.P., 4.A. 20,167 42.4 -

Material Handling, Equipment 16,323 23.9 l; Maintenance, and Miscellaneous Construction Activities Project Supervision and 10,000 12.9 Admi ni stration j O SUBTOTAL PHASE 11 141,680 4 21 .7 l

f 2614Q:1 ,

i f

Page 6 of 8 TABLE 6-2 ESTIMATE OF PERSONNEL RADIATION EXPOSURES FOR STEAM GENERATOR REPLACEMENT OPERATIONS AT POINT BEACH UNIT-I PHASE III: INSTALLATION ACTIVITIES O ESTIMATED ESTIMATED

, TASK LABOR EXPOSURE l DESCRI PTION (MN-HOURS) (MN-REM)

Steam Generator Lower Shell 7,744 12.7 Installation Installation of Reactor 43,666 193.1 Coolant Piping Steam Generator Girth Weld 19,135 9.9 Installation of Main Steam Piping 6,6 61 6.9 Installation of Feedwater Piping 4,715 2.3 Installation of Blowdown and 12,252 45.2 Miscellaneous Piping i

Install Steam Generator 7,622 9.6 i Level Instruments Installation of Insulation 7,747 25.0 Temporary Scaffolding 12,148 35.6 Temporary Lighting & Power 6,802 6.5 Cleanup and Decon 79,563 127.7 Polar Crane Operator 5,245 2.1 H.P., Q.A. 73,061 66.2 Material Handling, Equipment 35,777 25.5 Maint. , and Mi sc. Construction Activities Project Supervision & Administration 12,000 37.5 SUBTOTAL PHASE III 334,138 605.8 2614Q:1

Page 7 of 8 TABLE 6-2 ESTIMATE OF PERSONNEL RADI ATION EXPOSURES FOR STEAM GENERATOR REPLACEMENT OPERATIONS AT POINT BEACH UNIT-I PHASE-IV: POST INSTALLATION AND STARTUP ACTIVITIES j O

U  :

ESTIMATED ESTIMATED TASK LABOR EXPOSURE DESCRI PTION (MAN-HOURS) (Mall-REM) l Install Biological Shield Wall 2,121 2.9 Repair Crane Wall Opening 214 0.4 Install S/G Recin:ulation & Transfer 16,534 33.5  !

I System Remove Polar Crane Gibpol.e Mod. 1,500 5.0 Install Reactor Cavity Coaming 650 0.7 Reassemble Manipulator Crane 1,256 1.4 Remove S/d Transport System 200 1.5 Hydrostatic Tests 2,376 3.4 l

Temporary Scaffolding 3,382 6.4 Temporary Lighting & Power 1,71 2 1.5 Cleanup and Decon 14,378 23.2 Polar Crane Operator 1,186 0.5 t

Painti ng 9,000 8.0 H.P.,Q.A. 14,321 9.8 Mi scellaneous 3,000 5.0 l

l Material Handling, Equipment Maint., 10,000 9.7 and Miscellaneous Const. Activities Project Supervision & Administration 5,870 5.4 SUBTOTAL PHASE IV 87,700 118.3 i

sy%mkSt

Page 8 of 8 TABLE 6-2 ESTIETE OF PERSONNEL RADIATION EXPOSURES FOR l

STEM GENERATOR REPLACEMENT OPERATIONS i i

,' AT POINT BEACH UNIT-1 PHASE-V: STEAM GENERATOR STORAGE ACTIVITIES  ;

i  !

O ESTIMATED ESTIMATED l

! TASK LABOR EXPOSURE l

! DESCRIPTION (MAN-HOURS) (MAN-REM) l

! l,

.i Steam Generator 1,532 6.6 l Storage Activities  !

l l

i t

r l

1

, I i  !

I i

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O i O l 2614Q:1 l

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i l

}

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

i 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 confomance with local, state and federal regulations. I t

7.2 RESOURCES COMMITTED f

r

7. 2.1 Non-Recyclable Building Materials  ;

i I

Housekeeping operations for all construction areas will be perfomed 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.

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

! materials properly disposed.  ;

l 7.2.2 Land Resources 1 The repair effort will have minimal impact on existing site layout and j plant facilities. Two new facilities, a 16,000-square-foot operations

building and a 5,500-square-foot, temporary stean generator storage j building, will be constructed in the previously modified area adjacent  ;

to the plant to the north. A 6,500-square-foot access structure will be {

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constructed adjacent to the Unit I containment facade. Only about four

) acres of land will be required for construction activities. A parking lot will require about two acres and a layout area will require about 1.5 acres. The temporary steam generator storage building and opera-tions building will require about 0.5 acres.

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

Erosion and runoff control measures include: 1) limiting site grading and surface disturbance to the minimum area practicable, and 2) covering construction roads, parking, and laydown areas with gravel. In addi-tion, an approximately 300-foot-wide vegetated area will serve to filter sediments from any parking and laydown area runoff before reaching Lake Michigan. Following completion of the construction and repair activi-ties, remaining disturbed areas around the buildings will be seeded to return to grass cover.

pg The replacement steam generator lower assemblies will be deliv'ered to V the plant site by barge or rail. If the lower assemblies are delivered by barge, reconstruction of the temporary barge slip previously used for off-loading steam generators during initial plant construction will be required. The barge slip would be located approximately 2,400 feet south of the plant discharge structure. About 1,000 cubic feet of material would be excavated from the Lake Michigan shore for the ' slip and about 1,000 cubic feet of material would be dredged from the lake bottom for an approach channel. The bottom elevation of the barge slip and approach channel would be approximately 8 feet below low water datum for Lake Michigan. The excavated and dredged spoils would be placed in the existing land fill area on site. The barge slip and approach chan-nel would be allowed to revert to their original conditions upon comple-tion of the celivery operations or could be used as a recreational V boat-launching facility.

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i It is expected that the recreational boat-launch area and fisheman parking lot area, both located near the barge slip, would be temporarily closed to the public during construction activities and during steam generator delivery operations. The fisherman parking lot would be used for barge slip construction crew parking.

O The construction area to the north of the plant and the barge slip area  ;

are utilized minimally by the species of f auna known to inhabit the  !

plant site. It i3 anticipated that whatever species nomally use the habitat in the construction area will move to and use other adjacent O

t areas of the plant site once construction activities begin, and other existing populations of wildlife on the plant site will avoid the area until construction activities are completed. Approximately four acres will be lost. .

No displacement 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.

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No impact on rare and endangered species is expected. t 7.2.3 Water Resources A well to provide a permanent potable water supply for the existing -

construction building and the operations building is expected to be drilled in the area north of the plant. The well will have a capacity j of less than 70 gallons per minute. An estimated 5,000 to 10,000 gal-V lons of water per day will be used during construction. No impact on the existing groundwater aquifier is anticipated.

No groundwater impacts due to construction activities are anticipated.

l O No dewatering of the site is required. Holding tanks will be used for  ;

handling sanitary f acilities' wastewater and no groundwater discharges I will occur. t O ,

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Approximately 0.4 acres of lake bottom would be temporarily disturbed.

Turbidity would occur in an area slightly larger in size. Periphyton comunities, plankton populations, and benthic organisms in the area would be temporarily impacted. The barge slip area is not uniquely utilized by the few minnow species and slimy sculpin known to inhabit the shorezone adjacent to the plant site. It is anticipated that what-O ever fish species normally use or may occasionally frequent the barge slip area would move to other areas of the shorezone once construction activities begin and would avoid the area until construction is com-pleted. Similar shorline environment exists for several miles north and O

l south of the site and any impacts on the total aquatic environment of the site area would be minimal.

L 7.3 WASTE WATER 7.3.1 Sanitary Facilities L

Sanitary wastes will not be discharged on site. During site preparation and early stages of construction, port'able sanitary facilities will be d utilized. Wastes from these facilities will be removed by a licensed contractor. A holding tank will be installed to collect sanitary wastes i from the existing construction building and the operations building. A second holding tank will be installed to collect sanitary wastes from the containment access building. Wastes from the holding tanks will  !

also be removed by a licensed contractor. The holding tank for the l existing construction building and operations building will remain after construction is completed. The holding tank for the containment access -

building will be removed if the access structure is removed. Should the  ;

containment access building be left in place, the access building sani-tary f acilities may be connected to the plant sewage system. The 1 existing plant sewage treatment system has adequate capacity to process I

the additional wasteloads without modification.

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7.3.2 Laundering Operations O Laundry waste water generated during the repair activities will origi-nate from f acilities adapted for the proposed effort. If required, laundry waste water will be directed to the liquid radwaste system for processing (see Subsection 3.3.6.3). Additional information on the expected quality and quantity of laundry waste water from the steam generator repair program is provided in Subsection 5.2.2.4 of this report.

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

7.4.1 Noise Noise levels in the const'ruction area will be typical of those asso-ciated with the operation of site clearing equipment such as tractors, 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-ment, will be used to reduce noise levels. Occupational Safety and Health Administration Standards (OSHA) will be followed to protect personnel located on site. Noise inputs are expected to be confined to the construction area.

7.4.2 Dust O Dust will be created during site grading and by movement of vehicles on the unpaved construction areas. The primary means of control will be periodic sprinkling of the unpaved areas using water sprinkler trucks if (3 the need is indicated by visual inspection. The temporary parking lot U and laydown areas will be covered with stone or gravel. Only about 4 2600Q:1 7-5

acres will require grading. Following covering of the parking lot and laydown areas, only 0.5 acres of land for the operations building and l s steam generator storage building will remain in a disturbed state. The building sites will be open for a short period of time until the founda-tions are poured. Dust is not expected to be a problem and any minor impacts will be confined to the irvaediate areas near where the site O surface is disturbed.

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 is shown in Section 5.2.2. The radioactive effluent release points during steam generator repair activities will be the same as during O nomal plant operations.

Since releases of radioactive effluents during the repair program will be a small fraction of nomal operating plant releases and their potential 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 environmental monitoring program.

7.6 RETURN TO OPERATION 7.6.1 Water Use A well may be drilled north of the plant to provide a potable water source for the existing construction building and operations building.

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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7.6.2 Operational Exposure According to historical data on occupational radiation exposure for the Point Beach Nuclear Plant, presented in NUREG-0713, " Occupational Radia-tion Exposure at Commercial Nuclear Power Reactors", the average annual exposure per unit during 1979,1980, and 1981 was 306 man-rem. During V the preceeding six years, this annual average per unit was only 205 man-rem. The additional radiation exposure in recent years is primarily due to increased steam generator inspection and maintenance activities.

Assuming that the replacement steam generator tubes maintain their inte-grity during the remaining operating lifetime of the plant (27 years),

the radiation exposure should be reduced by at least this incremental amount (approximately 100 man-rem per year). Factoring in the estimated radiation exposure expected for the repair effort, given in Table 6.6-1 (about 1400 man-rem), a total of 1300 man-rem may be saved over the lifetime of the plant by repairing the steam generators.

7.6.3 Radiological Releases 1 Secondary plant releases result from primary to secondary leakage.

While doses due to radioactive releases from the existing steam generators are insignificant, the repaired steam generators will have enhanced tube integrity thus further reducing doses due to secondary plant releases.

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