ML17263A831

Draft Rev C to Design Criteria Ginna Station Containment Structural Mods Wbs 4

ML17263A831
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Site:	Ginna
Issue date:	09/26/1994
From:	BECHTEL CORP.
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ML17263A830	List: ML17263A829 ML17263A831 ML17309A558
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DC-10034A-DRFT, DC-10034A-DRFT-RC, NUDOCS 9411080319
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QEbK ~:QECHTEL'CORPORATION i 9-26-94 t

15:27 'ECHTEL SGR GROUP~

,h 17167716828'0 4(t' Design Criteria Gmna Stat<on CONTAINMENTSTRUCTORAL MODIFlCATIONS WBS 4 Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14549 DC-1'0034A EWR 10034 Revision C September 26, 1994 Prepared By:

Reviewed By:

Reviewed By:

Reviewed By:

Approved By:

Reviewed By:

Reviewed By'.

Approved By:

Preparer/Bechtel Mechanical/Nuclear Supervisor/Bechtel Civil Supervisor/Bechtel, Electrical/Control Systems Supervisor/Bechtel Project Engineer/Bechtel Responsible Engineer(s)/RGAB Quality Assurance/RGEcB DATF DATE DATH DATE DATE DATE Manager, Steam Generator Replacement Group/RG&E DATE Page J,of 97 4411080319 941101 PDR ADQCK 05000244 P

PDR

SENT BY:BECIffEL CORPORATION l 9-26-94 l

15:28 'ECHTEL SGR GROUP~

17167716828;~ 5/ 6 e isi n Page 1

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h Page 85 Latest Revision Latest Revision Latest Revision 87 88 91 92 95 96 97 Attachment 1

C Appendix A (l2 pages)

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DESIGN CRITERIA INTRODUCTION Design Criteria Documents for the various engineering and construction activities associated with the Steam Generator Replacement Project are as follows:

Design Criteria Document Bechtel WBS Title DC-10034 Replacement Steam Generator Procurement DC-10034A DC-10034B DC-10034C WBS 4

WBS 5

WBS 6

WBS 7

WBS 8

Containment Structural Modifications Steam Generator Rigging s Handling Steam Generator Vessel Steam Generator Piping Insulation DC-10034D WBS 10 Temporary Utilities s Services DC-10034E WBS 3

Facilities Outside Containment DC-10034F WBS 11 Testing 8 Inspection WBS:

Work Breakdown Structure The purpose of this Design Criteria document is to define the requirements for engineering designs and physical construction activities associated with Bechtel WBS Group 4, "Containment Structural Modifications."

The WBS group, and subsequently this document, is divided into general categories.

This Design Criteria.

has been divided such that each of the categories has been addressed in its entirety, independent of the other categories.

The Design Criteria is formatted as fo13.ows:

SECTION-CATEGORY 1.0 2.0 3.0 4.0 5.0 6.0 Containment Construction Openings Temporary Enclosures and Laydown Area Pexmanent Steam Generator Supports Steam Generator Lower Support Temporary Restraints Reactor Cavity Decking Interfering Commodities Design Criteria DC-10034A EWR 10034 Page 4

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SECTlON CATEGORY 7.0 New Steam Generator Secondary Manway Access Platforms Attachment 1

10CFR50.49 (EQ) Applicability Form Appendix A Bechtel Data Letter No.

88, "Concrete Anchorage for New Anchor Bolts," dated November 30, 1993.

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+ inches.

The layers are oriented such that the bars in each layer make a 120-degree angle with the bars in an adjacent layer.

1.1.1.3.3 The thickness of the cylindrical shell is 3Y> feet and the thickness of the dome is 2'eet.

The cylinder and dome liners are %-inch-thick.

The inside radii of the cylinder and the dome are 52'> feet.

The cylinder shell is 99 feet high (from its lower edge at El. 231'-8" to the spring line at El. 330'-8").

The upper anchors of the vertical tendons are on the dome at El. 343'-2" Design Criteria DC-10034A EWR 10034 Page 6

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0 (12.5 feet above the spring line).

The hei containment shell is 154 feet.

The base of the containment is a 2-foot-thick concrete slab.

The base slab liner is /4-inch-thick.

Xn addition, a 2-foot-thick concrete fillis superimposed on the base slab liner.

The top of the base slab liner is at an elevation slightly above El. 233'-0" and the grade line is at El. 270'-0".

1.1. 1.4 Construction Sequence The primary construction tools that will be used to create the openings in the dome reinforced concrete will include automated hydraulic jackhammers, manual jackhammers, arid rock. drills.

These tools will be used to chip the concrete off of the reinforcing steel.

As the rebar is exposed, it will be cut into manageable lengths to facilitate removal.

Following removal of the reinforced concrete, the steel liner plate will be cut in sections, lifted from the containment

dome, and moved to a preparation area onsite to prepare it-for re-installation.

Restoration of the containment dome will involve re-installing the cut sections of liner plate, cadwelding or stick-welding new rebar in place, and placing new concrete.

The following is a description of the planned construction sequence to create and restore the construction openings in the containment dome.

The construction sequence is divided into 5 stages and provides a reference basis for the containment dome structural evaluation approach.

1.1.1.4.1 Stage 1

Prior to shutdown for the steam generator replacement (SGR) outage with the plant in an Operating

Node, the following construction activities will be accomplished:

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Temporary structural steel platforms or frames will be installed to support the automated hydraulic jackhammers, temporary service cranes, craft personnel, and miscellaneous construction equipment and materials.

See Section 2.0.

Preliminary concrete scoring and rock drilling operations will commence.

These operations will be limited to approximately the top 8 inches of concrete on the dome.

No automated hydraulic jackhammers will be used during these operations.

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1.1.1.4.2 Stage 2

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With the plant in a Cold Shutdown or Refueling Condo.tron, with fuel in the reactor vessel and/or in the process of being offloaded to the spent fuel pool, the following construction activities will be accomplished:

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Concrete excavation operations will commence using the automated hydraulic jackhammers, manual jackhammers, and rock drills.

As the rebar is exposed, it will be cut into manageable lengths to facilitate removal.

Chipping operations will continue and will be completed down to the top of the liner plate.. During this plant condition, the liner plate will not be penetrated and its leaktight integrity will be maintained.

Xnterferences on the inside of the dome will be temporarily removed (containment spray piping, HVAC ductwork, and painter's trolley).

See Section 6.0.

1.1.1.4.3 Stage 3

With the plant in a "no-mode" condition with all fuel offloaded to the spent fuel pool, the reactor cavity

drained, and all necessary isolation of the spent fuel pool and supporting containment systems

achieved, the following construction activities will be accomplished:

Concrete excavation operations may still be on-going.

Once completed; the steel liner plate will be cut in sections, lifted from the containment

dome, and moved to a preparation area onsite to prepare it for re-installation.

The old. steam generators will be removed and the new steam generators will be installed.

The liner plate sections will be prepared, rigged into place,

welded, NDE performed, and leak-tested.

Stiffeners and leak chase channels will be installed as necessary.

The interferences on the inside of the dome will be reinstalled and tested as appropriate.

The majority of the reinforcing bars will be installed and cadwelded.

Splicing of rebar by welding may also be per formed, if necessary.

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1.1.1.4.4

,Stage 4

With the plant returned to a Refueling or Cold Shutdown Condition, with fuel in the reactor vessel and/or in the process of being onloaded from the spent fuel pool, the following construction activities will be accomplished:

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The final pieces of reinforcing bar will be set and cadwelded/stick-welded.

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All concrete will be placed and cured.

Strength tests will be conducted and acceptable results will be achieved.

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A limited containment structural integrity test (SIT) will be conducted (see Section 1.24)

and, based on acceptable test results, full containment integrity will be proven.

1.1.1.4.5 Stage 5

With the plant returned to an Operating Condition, the following construction activities will be accomplished:

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The temporary structural steel platforms, automated hydraulic jackhammers, temporary service

cranes, and miscellaneous construction equipment will be removed.

1.1.1.5 Rigging and handling of heavy loads, including the steam generators, liner plate sections, and construction equipment to be mounted on the containment dome is addressed in Design Criteria Document DC-10034B.

1.1.2 1.1.2.1 Functions Consistent with the planned construction sequence outlined in Section 1.1.1, basic functions are as follows:

1.1.2.1.1 Operating Condition The containment vessel is a safety-related, seismic Category I structure which must perform the following functions:

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Sustain without undue risk to the health and safety of the public the initial effects of gross equipment failures, such as a large reactor coolant pipe break, without loss of required integrity.

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Retain for as long as the situation requires the functional capability of the containment to the extent necessary to avoid undue risk to the health and safety of the public.

The containment liner plate provides a leaktight seal for the containment vessel.

The liner plate is an integral part of the containment structural system.

1.1.2.1.2 Cold Shutdown or Refueling Conditions During cold shutdown or refueling conditions, the safety-related function of the containment dome concrete and liner plate is limited to providing a leaktight barrier against the release of radioactivity.

The containment dome also serves a seismic Category II/I function to ensure that it does not fail and impact the fuel or adjacent safety-related structures during a postulated seismic event.

During the offload and onload of fuel to the reactor

vessel, the containment liner plate also serves the safety significant, seismic Category II/I function of acting as a form for the removal and placement of the reinforced concrete.

1.1.2.1.3 "No Mode" Operation During "no mode" operation with the reactor vessel completely defueled, the reactor cavity drained, and all necessary isolation of the spent fuel pool and supporting containment systems

achieved, the containment performs the safety significant, seismic Category II/I function of ensuring that it does not fail and adversely impact the spent fuel pool or other adjacent safety-related structures.

The non-safety functions of the concrete and liner plate during this period are:

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To support construction equipment and personnel during the steam generator replacement outage.

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To,support the snow and wind loads applied to the dome.

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To act as a form for the removal and placement of the reinforced concrete.

1.1.3 1.1.3.1 Performance Requirements The construction activities during the outage shall not impair the ability of the containment dome to perform its required seismic Category I functions.

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1. 1.3.2 For the steam generator replacement, the co crete containment dome shall provide structural support for the functions described in Section 1.1.2 in the event of wind,

snow, and seismic conditions, as applicable.

1.1.3.3 The concrete and liner plate shall withstand the external forces necessary to remove and reinstall the reinforced concrete without adversely affecting the surrounding dome concrete or the ability of the dome to perform its designated functions.

1.1. 3.4 The performance requirements of the containment dome after the openings are created are to support its own deadweight and construction loads and to ensure that it does not fail and impact adjacent safety-related structures during a postulated seismic event.

1.1.3.5 The performance requirements of the containment dome after the steam generator replacement are to maintain its structural capacity in accordance with its safety-

related, seismic Category I design.

1.1.4 1.1.4.1 Control The removal and re-installation of the containment dome

concrete, reinforcing bar, and liner plate shall. not impair the structural capacity of the containment vessel.

Therefore, there is no requirement for additional controls on any, existing systems,

features, or equipment.

1. 1.5 1.1.5.1 Modes of Operation The containment dome shall be capable of performing its required functions under all modes of plant operation.

See the discussion in Sections 1.1.1 and 1.1.2.

1.2.0 1.2.1 Referenced Documents Bechtel Quality Assurance Program Plan for Rochester Gas

& Electric Corporation, R.

E. Ginna Nuclear Power

Plant, Steam Generator Replacement

Project, Bechtel Job No.

22225.

1.2.2 Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

1.2.3 R.

E. Ginna Nuclear Power Plant Steam Generator Replacement Project

- Containment Opening Report, July 1993, Bechtel.

1.2.4 Rochester Gas and Electric, Ginna Station Unit No.

1, Containment Vessel Analysis, Volumes 1,

2, 3,

1968, Work Order 4155.

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1.2.5 Special Processes

Manual, Ginna Nuclear Power lant Steam Generator Replacement

Project, Bechtel Job No.

22225.

1.2.6 1.2.6.1 1.2.6.2 1.2.7 1.2.7.1 1.2.8 1.2.8.1 1.2.8.2 1.2.8.3 1.2.8.4

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1.2.8.5 1.2.8.6 American Concrete Institute (ACI)

ACI 318-63, Building Code Requirements for Reinforced Concrete and Commentary.

ACI 301-66, Specifications for Structural Concrete for Buildings.

I'merican Institute of Steel Construction (AISC)

Manual of Steel Construction, 9th edition.

American Society of Mechanical Engineers (ASME)

ASME Section II, 1986 Edition, SA-370, Specification for Mechanical Testing of Steel Products.

ASME Section III, 1965 (Provisions of Article 4 as defined in UFSAR Section 3.8.2.1.1.2).

ASME Section III, Rules for Construction of Nuclear Power Plant Components, 1986 Edition.

ASME Section III, Division 2, Subsection CB, Subparagraph CB-4333.2, 1986 Edition.

ASME Section VIII (undated),

(Fabrication Practices for Welded Vessels Only).

ASME/ANSI N45.2.5-1974, Supplementary Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete, Structural Steel,

Soils, and Foundations During the Construction Phase of Nuclear Power Plants.

1.2.9 1.2.9.1 American Society for Testing and Materials (ASTM)

ASTM A36, Revision A-93, Standard Specification for Structural Steel.

1.2.9.2 1.2.9.3 1.2.9.4 ASTM A185, Revision A-90, Standard Specification for Welded Steel Wire Fabric for Concrete Reinforcement.

ASTM A307, Revision A-93, Standard Specification for Carbon Steel Bolts and Studs 60,000 psi Tensile Strength.

ASTM A325-93, Standard Specification for Structural Bolts, Steel, Heat Treated 120/105 ksi Minimum Tensile Strength.

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1.2.9.5 1.2.9.6 ASTM A442-60T, Tentative Specification for Carbon Steel Plates with Improved Transition Properties.

ASTM A408-64T, Specification for Special Large Size Deformed Billet-Steel Bars for Concrete Reinforcement.

1.2.9.7 ASTM A490-93, Standard Specification for Heat-Treated Steel Structural Bolts, 150 ksi Tensile Strength.

1.2.9.8 1.2.9.9 1.2.9.10 ASTM A516-90, Standard Specification for Pressure Vessel Plates, Carbon Steel, for Moderate-and Lower-Temperature Service.

ASTM A563-93, Standard Specification for Carbon and Alloy Steel Nuts.

ASTM A615, Revision B-92, Standard Specification for Deformed and Plain Billet-Steel Bars for Concrete Reinforcement.

1.2.9.11 1.2.10 1.2.10.1 1.2.10.2 1.2.11 1.2.11.1 ASTM F436-93, Specification for Hardened Steel Washers.

American Welding Society, (AWS)

AWS D1.1-94, Structural Welding Code.

AWS D1.4-92, Structural Welding Code-Reinforcing Steel.

American Society of Civil Engineers (ASCE).

ASCE 7-93, Minimum Design Loads for Buildings and Other Structures.

1

~ 2. 12 1.2.12.1 1.2.13 1.2.13.1 1.2.13.2 Concrete Reinforcing Steel Institute (CRSI)

CRSI MSP-1, 25th Edition, Manual of Standard Practice.

Bechtel Design Guides EDG-C0313, Revision 1, "Civil/Structural Engineering Design Guide for Concrete Expansion Anchors."

EDG-C0102, Revision 1,

"Thermal Effects Combined With Real Loads."

1. 2. 14

1. 2. 15 "Bechtel Structural Analysis Program (BSAP)

User's/Theoretical Manual",

Volumes I

& II, Revision 18, Bechtel Power Corporation, December 17, 1993.

Winter, G.,

and Nilson, A. H., "Design of Concrete Structures,"

9th Edition, McGraw-Hill Book Company, New

York, New York, 1979.

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1.2.16 NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"

Revision 1, July 1981.

1. 2. 17 Timoshenko, S.,

and Woinowsky-Krieger, S.,

"Theory of Plates and Shells,"

2nd Edition, McGraw-Hill Book Company, 1959.

1. 2. 18 NRC Safety Guide 1.18, Structural Acceptance Test for Primary Reactor Containments.

1. 2. 19 Official Compilation of Code's, Rules and Regulations of the State of New York, Department of State,

Albany, New York, December 1986 (Copyright 1961).

1. 2. 20 Design Criteria Document for WBS 5, Steam Generator Rigging s Handling, DC-10034B.

1.2.21 RGEE Specification CE-125, Technical Requirements for Furnishing and Erecting Structural Steel, Revision 7.

1.3.0 1.3.1 Seismic Cate or The containment vessel dome and liner are classified as safety-related, seismic Category I structures.

1.3.2 During the "no-mode" phase of operation with all fuel offloaded from the reactor vessel, the reactor cavity

drained, and all necessary isolation of the spent fuel pool and supporting containment systems

achieved, the containment vessel and liner serve no safety functions.

However, the containment vessel and liner serve a

safety significant, seismic Category II/I function to ensure they do not fail and adversely impact the spent fuel pool or adjacent safety-related structures during a seismic event.

1.3.3 All activities associated with the physical removal and reinstallation of the containment dome concrete, reinforcing bar, and liner plate are classified as safety-related to ensure there. are no residual effects on the adjacent sections of the containment dome and that the restored containment dome satisfies all applicable engineering and construction requirements.

1.4.0 1.4.1 1.5.0 1.5.1 ualit Grou Code Class Fluid S stems Onl Not Applicable.

Electrical S stem Safet Classifications Not Applicable.

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1.6.0 1.6.1 1.7.0 A Pro ram A licabilit The Bechtel Quality Assurance Program Plan shall apply to the structural modifications and repairs associated with the containment construction openings.

Codes Standards and Re lator Re uirements 1.7.1 1.7.2 1.7.2.1 1.7.2.2 1.7. 3 1.7.3.1 1.7.3.2

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1.7.3.3 1.7.4 1.7.4.1 1.7.4.2 1.7.5 1.7.5.1 1.7.5.2 1.7.6 1.7.6.1 The following principal codes, standards, and regulatory requirements shall be used in the design and analysis of the containment openings:

Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

American Concrete Institute (ACI)

ACI 301-66, Specifications for Structural Concrete for Buildings.

~ ': rye, ACI 318-63, Building Code Requirements for Reinforced Concrete and Commentary.

American Institute of Steel Construction (AISC)

Manual of Steel Construction,'th edition Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings, adopted April 17, 1963.

Code of Standard Practice for Steel Buildings and,

Bridges, revised February 20, 1963.

American Society of Mechanical Engineers (ASME)

ASME B&PV Code,Section III, 1965 (Provisions of Article 4 as defined in UFSAR Section 3.8.2.1.1.2).

ASME B&PV Code,Section VIII (undated),

(Fabrication Practices for Welded Vessels Only).

American Welding Society (AWS)

D1.1-94, Structural Welding Code Steel.

AWS D1.4-92, Structural Welding Code-Reinforcing Steel.

American Society of Civil Engineers (ASCE)

ASCE 7-93, Minimum Design Loads for Buildings and Other Structures.

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Official Com ilation of Codes, Rules and Regulations of p

the State of New York, Department of State,

Albany, New York, December 1986 (Copyright 1961).

1.8.0 1.8.1 Desi n Conditions The containment dome shall be evaluated in accordance with its original design codes.

All operating design conditions shall remain unchanged.

The containment dome shall be capable of performing its required functions under all modes of plant operation.

See the discussion in Sections 1.1.1 and 1.1.2.

1.8.2 1.8.2.1 Concrete Shell The structural evaluation of the concrete shell of the dome will be based on ACI 318-63, Part IV-B, "Ultimate Strength Design."

1.8.3 1.8.3.1 1.8.3. 1. 1 Dome Liner System Existing Design The structural evaluation of the liner system of the dome will be based on the 1965 ASME BaPV Section III, supplemented by Sections II, and VIII. This is consistent with the original design criteria identified in UFSAR Section 3 '.2.1.1.2.

1.8.3.1.2 The design of the dome liner system was re-evaluated in great detail in UFSAR Section 3.8.2.3.

The analytical and test results from this re-evaluation, including the liner buckling strength and liner-stud interaction, will be used for the structural evaluation of the dome liner system attached to concrete.

Specifically, the liner system re-evaluation in UFSAR Section 3.8.2.3 concludes that while buckling of the liner plate may occur at below yield stress levels (elastic buckling),

the studs are capable of resisting the ensuing shear forces as long as the evaluated liner stress does not exceed the following limits:

(a) 29 ksi in Zone 1 which extends from the dome spring line up to a 50 arc above the spring line.

(b) 12 ksi in Zone 2 which extends from an arc beginning at 55'bove the spring line up to the dome apex.

The above allowable stress limits, along with a linear interpolation for determination of the allowable stress limit for the region (arc) between 50 and 55'bove the spring line, will be satisfied for evaluation of the liner system stiffened by concrete.

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ensure that a stud failure will not occur even if the surrounding liner panels were to buckle.

1.8.3.2 1.8.3.2.1 Construction Sequence For the bare liner regions exposed during the concrete chipping operations and during the placement of new

concrete, factors of safety of 1.5 and 1.1 will be used to determine the allowable buckling stress for normal and seismic Category II/I loading combinations, respectively.

In the event that the calculated bare liner compressive stresses exceed the buckling allowables, a suitable stiffening arrangement will be used to increase the effective liner thickness

and, hence, the corresponding allowables.

1.8.3.2.2 The design evaluation for the construction activities will also limit the maximum allowable stress to 0.75 Fy to ensure that the liner has adequate safety margin against yield stress and permanent deformation.

1.8.4 Finite Element Modelling Associated With The Construction Openings e

1.8.4.1 A finite element approach will be used for the structural analysis to demonstrate that the. containment shell structure is capable of adequately carrying the.

applicable loads during and after the dome opening construction and restoration.

The finite element, model used for this structural analysis is described below.

1.8.4.2 The two containment openings on.the dome will be far away (several times the dome shell thickness) from the spring line which is the transition zone between the cylindrical shell and the dome.-

Therefore, only the dome (hemisphere) will be directly included in the finite element model.

The finite element model will begin at the spring line and extend to the top of the dome.

The stiffness (or flexibility) of the cylindrical shell at the spring line will be represented by a set of nodal springs along the edge (spring line) of the dome in the finite element model.

The spring constants of the nodal springs along'he edge (spring line) of the dome will be analytically calculated based on the structural properties of the cylindrical shell from its base to its spring line.

In the calculation of these spring constants, the effects of the equipment hatch, personnel access

hatches, and all other penetrations will be neglected.

This is justified because all of these hatches and penetrations are at a remote distance from the spring line and,

thus, they have insignificant effects on the stiffness the cylindrical shell at the spring line.

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The advantages of the above model over a finite element model that would include the entire containment shell are:

(i) it is adequate and less

complex, and (ii) it keeps the focus of the evaluation only on the

dome, as appropriate.

1. 8.4. 3 Except in the regions where the liner will be bare (exposed),

the structural loads, as de'fined in the finite element model, will be resisted solely by the reinforced concrete.

This is consistent with standard practice and is in accordance with the original design.

The liner plate was intended only as a leaktight barrier and not as a load-resisting structural element.

Since the concrete shell is much stiffer than the liner

plate, the strains in. the liner will essentially follow those in the concrete.

Therefore, the structural evaluation of the liner plate attached to concrete will be based on imposing the concrete strains (obtained from the concrete shell finite element analysis) onto the liner plate.

The regions where the liner will be exposed will be modelled using bare liner properties, enabling a direct evaluation of liner stresses in these regions.

1.8.4.4 The CE-800 BSAP computer program will be used for the structural analysis of the containment dome.

In the finite element

model, the dome concrete shell will be idealized by a set of thin plate/shell elements (LCCT9s in the CE800 BSAP computer program).

This is appropriate because the dome concrete shell behaves as a thin shell structure for the following reason:

Based on the book "Theory of Plates and Shells,"

a thin shell approach is valid if the mean radius of curvature of a shell is several times the shell thickness.

For the dome concrete shell under consideration, the mean radius of curvature is 53.75 feet and the thickness is 2.5 feet; thus, the radius of curvature is 21.5 times its thickness.

Consequently, the dome concrete shell behaves as a thin shell structure.

For the structural

analysis, the thickness of the LCCT9 elements will be the concrete shell thickness of the dome consistent with:

(a) the current practice of reinforced concrete structural analysis for structural element proportioning, (b) other containment finite element models recently performed by Bechtel and subsequently reviewed and accepted by the

NRC, and (c) although not a design requirement for this containment, it is also consistent with Subsection II of SRP Section 3.8.1.

This means that the structural analysis of the concrete dome will be based on the uncracked condition.

1.8.4.5 As discussed in Section 1.8.4.4, the finite element analysis of the concrete dome will be based on the uncracked condition.

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bending moment on the dome shell, obtained from the finite element analysis, will correspond to the uncracked section.

For the cracked reinforced concrete

section, the thermal-induced bending moment will be considerably lower than the uncracked bending moment.

The reduced thermal bending moment due to section cracking will be calculated from the uncracked section thermal bending moment based on the approach provided in Bechtel Design Guide EDG-C0102.

1.8.4.6 In terms of nodal coordinates, element

numbers, and element connectivity, the finite element model of the containment dome, for each of the 5 stages of the construction sequence described in Sections 1.1.1.4.1 through 1.1.1.4.5, will be basically the same.

The finite element model associated with each stage of the construction sequence can be represented by changing the properties of the LCCT9 elements within the footprints of the two openings, as described in Examples 1 and 2 below.

EXAMPLE 1 Consider Stage 3 of the construction sequence with the concrete and liner removed to create the two openings.

Openings can be created in the finite element model by declaring all the

-nodes in the openings as fixed and eliminating the element connectivity in the opening regions.

EXAMPLE 2 Consider Stage 3 of the construction sequence with only the liner in the opening footprints attached to the containment, dome.

The bare liner can be represented in the finite element model by setting the thicknesses, Young's Moduli, Poisson's

Ratios, thermal expansion coefficients, and mass densities of all the elements in the opening footprint equal to those of the liner.

In this case, the neutral planes of the liner elements and the neutral planes of their surrounding concrete elements are concentric in the finite element model; however, in reality they are eccentric.

The effect of this eccentricity will be checked.

1. 8.4.7 Dead load of the dome is essentially the only permanent

'load.

The compressive stress in the concrete due to dead load on the dome is expected to be so small that the concrete creep and its structural effects on the new concrete, placed in the two opening regions, will be insignificant.

This will be verified by a calculation using the results of the finite element analysis associated with the containment shell dead load.

Consequently, the concrete creep and its effects will not be further considered in the finite element modelling and stress analysis of the containment dome.

Design Criteria DC-10034A EWR 10034 Page 19 Revision C

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A mix design for the new concrete will be created such that its mechanical properties (e.g.,

the specified cylinder strength, or f,', etc.)

are comparable with the existing dome concrete mechanical properties.

Therefore, in the finite element modelling, the mechanical properties of the new concrete will be considered to be the same as the mechanical properties of the existing concrete.

1.8.4.9 For the dome restoration, either the existing liner

system, a comparable new liner system (e.g.,

ASTM A516 Grade 60 liner plate is equivalent to the existing ASTM A442-60T Grade 60 liner plate),

or a combination of new and existing will be used.

For anchoring, studs and/or stiffeners will be used.

Therefore, for the structural

analysis, the restored liner plate and anchoring system is considered to be equal or better than the existing one.

1.8.5 1.8.5.1 Finite Element Model Validation'he finite element model for the containment dome shall be validated by Method 1 and/or Method 2 described below:

1.8.5.1.1 Method 1.

This validation shall be based on the original containment shell model prior to the start of any construction opening activities.

The validation shall be performed by:

(a) comparing the dead load membrane

forces, shears, and bending moments in the

dome, obtained from the finite element model described in Section 1.8.4, with the dead load membrane

forces, shears, and bending moments in the dome from the original analysis, available in the original calculations, and (b) comparing the pressure membrane

forces, shears, and bending moments in the

dome, obtained from the finite element model described in Section 1.8.4, with the pressure membrane

forces, shears, and bending moments in the dome from the original analysis, available in the original calculations.

1.8.5.1.2 Method 2

ln the same manner as Method 1, this validation shall be based on the original containment shell model.

The validation shall be performed by:

(a) comparing the dead load membrane

forces, shears, and bending moments in the dome, obtained from the finite element model described in Section 1.8.4, with the dead load membrane

forces, shears, and bending moments in the dome obtained from a classical analytical solution, and (b)

Design Criteria DC-10034A EWR 10034 Page 20 Revision C

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comparing the pressure membrane

forces, shears, and bending moments in the

dome, obtained from the finite element model described in Section 1.8.4, with the pressure membrane

forces, shears, and bending moments in the dome obtained from a classical analytical solution.

1.8.5.2 Both Method 1 and Method 2 include a pressure load case.

Since the finite element model is limited to the dome section, it does not have the capability to accommodate load transfer from the cylindrical shell.

Therefore, under the pressure

loading, there may be a

slight underestimation of the hoop membrane and meridional bending stresses near the spring line.

Despite this underestimation of membrane forces and moments at the spring line, the pressure load case is an adequate validation method because the prediction of forces and moments at distances a few times the dome thickness above the spring line (the region of interest) remains accurate.

Furthermore, the load combinations for Stages 2,

3, and 4 of the construction sequence will not include a postulated accident pressure load since the plant will not be in operation.

Therefore, the underestimation of hoop membrane forces and meridional bending moments for the pressure loading case have little significance as far as the actual loading during the construction activities is concerned.

Load Conditions 1.9.1 Dome Structure Evaluation for Stage 1 of the Construction Sequence 1

~ 9.1.1 For Stage 1,

no new analysis or, evaluation of the dome is required because the original structural analysis and evaluation of the containment dome remain valid for the following two reasons:

1.9. 1. 1. 1 As described in UFSAR Section 3.8.1.2.3.1, the containment shell was designed based on the following three load combinations (LC1S1,

LC2S1, and LC3S1):

LC1S1

= 0.95 D + 1.5 P

+ 1.0 T + 1.0 Fv LC2S1

= 0.95 D + 1.25 P

+ 1.0 T' 1.25 (E or W)

+ 1.0 Fv LC3S1

= 0.95 D + 1.0 P

+ 1.0 T + 1.0 E' 1.0 Fv WHERE:

D =

Dead weight of structure Design Criteria DC-10034A EWR 10034 Page 21 Revision C

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26 94

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P

=

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T, T',

T =

Thermal loads described in UFSAR Section 3.8.1.2.3.3 E

=

OBE (Operating Basis Earthquake) load E'=

SSE (Safe Shutdown Earthquake) load W =

Wind load as described in UFSAR Section 3.3.2.1.6 Fv Final vertical prestress force.

Fv is applied at 12.5 feet above the spring line on the dome.

Per the original calculation, Fv = 299 k/ft which is also consistent with the UFSAR description on Pg.

3.8-33.

In the

UFSAR, 1.0 Fv is not explicitly included in LC1S1,

LC2S1, and LC3S1.

Since Fv is always present, it is explicitly included, in the above load combinations for completeness.

The loading effects of installing the temporary structures and equipment on the containment dome with respect to LC1S1,

LC2S2, and LC3S3 are expected to be insignificant because:

(i) the mass (or weight) of the temporary components is much less than the mass of the containment shell which will lead to only a slight increase in dead load and seismic load effects; (ii) the accident pressure (P) in LC1S1,

LC2S1, and LC3S1 is the preponderant load; and (iii),the dead weight of the temporary components actually acts against the accident pressure

and, thus, reduces the load effects of the accident pressure.

Nevertheless, an evaluation will be provided to show that the load effects of the temporary components are indeed insignificant.

1.9.1.1.2 Based on the moment diagrams of the dome for load combinations

LC1S1, LC2S1, and LC3S1, obtained from the original analysis as shown in UFSAR Figures 3.8-13 through 3.8-15, there is essentially no bending moment in the dome even in the region near the spring line.

This means that the entire dome essentially behaves as a pure membrane shell.

In all three load combinations, the membrane forces are all in tension with a minimum magnitude of 204 k/ft (kips per foot).

For the dome concrete with f,'f 5000 psi (UFSAR Section

3. 8. 1. 6. 1. 1), its membrane (true) tensile strength is at best equal to 353 psi (5 square root f,'er the book "Design of Concrete Structures" ).

The membrane tensile Design Criteria DC-10034A EWR 10034 Page 22 Revision C

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strength of the dome concrete is at most 1

k ft (0.353 x 2.5 x 1.0 x 144).

Therefore, under LClS1, LC2S1 or LC3S1, the dome concrete will be cracked and the rebar will carry all the membrane forces.

Consequently, in Stage 1 of the construction

sequence, even though approximately the first 8 inches of concrete in th'e opening footprints is cut/chipped, the dome structural capability to sustain

LC1S1, LC2S1, or LC3S1 remains the same.

1.9.2 Dome Structure Evaluation for Stage 2 of the Construction Sequence 1.9.2.1 In Stage 2 of the construction

sequence, the containment dome structure, with the openings partially or fully chipped out,. must be evaluated as a seismic Category II/I structure for the following load combination (LC1S2):

LC1S2

=

1.0 D + 1.0 P

+ 1.0 T

+ 1.0 E' 1.0 Fv Where:

P Accident pressure load associated with cold shutdown or defueling or refueling condition.

P

=

0 for the Stage 2

condition during the SGR outage.

For undisturbed concrete

elements, use the thermal load associated with the dome temperature gradient shown in UFSAR Figure 3.8-8.

For disturbed concrete elements (if applicable) and bare (exposed) liner plate elements, the temperature loadings will be considered as the maximum and minimum of the range of temperatures possible due to factors such as inside and outside temperature, heat. transfer coefficient, etc.

E' SSE load, as applicable for the dome configuration in Stage

2. E'ill be expressed as appropriate equivalent static forces in all three directions for regions affected by the construction activities.

For the undisturbed remainder of the

dome, the equivalent static forces will be determined based on UFSAR Section 3.7.2.1.1.

1.9.2.2 Since live load is not included in LC1S2, the gravity load or weight due to the temporary platforms and construction equipment will be included in LC1S2 as dead load (D).

Design Criteria DC-10034A EWR 10034 Page 23 Revision C

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1. 9' 2. 3 All the load factors are unity in load combination LC1S2 because, at this stage, the containment serves a

seismic Category II/I function.

For the structural adequacy evaluation of the dome as seismic Category II/I, (a) the allowable capacity of the dome concrete shell will be taken as the nominal (minimum) ultimate strength per ACI 318-63, Part IV-B; and (b) the allowable capacity of the dome liner system will be taken as 0.9 of its nominal yield strength.

In

addition, the limiting liner (compressive) stresses per Section 1.8 '

will be satisfied.

1.9.2.4 In addition to the seismic Category II/I evaluation described

above, the dome structure at this stage must also be evaluated for.its normal loading condition.

This evaluation will be based on the following two sets of load combinations:

SET 1:

. Load Combinations (LC1S2C,

LC2S2C, and LC3S2C) for the Dome Concrete Shell LC1S2C

= 1.5 (D + To)

+ 1.8 L + 1.0 Fv LC2S2C

= 1.25 (D + To + L + Sn

+ W)

+ 1.0 Fv LC3S2C

= 0.9 (D + To)

+ 1.1 (Sn

+

W)

+ 1.0 Fv Where:

L = live load Sn

= Snow load as specified in UFSAR Section 3.8.1.2.2 W = Wind load as described in UFSAR Section 3.8.1.2.2 I

To = Operating thermal load described in UFSAR Section 3.8.1.2.3.3

LC1S2C, LC2S2C, and LC3S2C are based on the required load combinations specified in ACI 318-63, PART IV-B.

In the State of New York Building Construction

Code, seismic load is not required to be considered in the building design.

Consequently, seismic load is not included in load combinations LC2S2C and LC3S2C above and, likewise, load combination LC1S2L below.

Design Criteria DC-10034A EWR 10034 Page 24 Revision C

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SET 2:

Load Combination

{LC1S2L) for the Liner System LC1S2L ="D + L + Sn

+

W + To

+ Fv 1.9.2.5 The temporary platforms and construction equipment will be included as live load in LC1S2C,

LC2S2C, LC3S2C, and LC1S2L.

1.9.2.6 1.9.3 For all the load combinations in Sets 1 and 2, the structural evaluation of the concrete shell and liner system of the containment will be performed based on the structural codes and guidelines identified in Sections 1.8.2 and 1.8.3.

Dome Structure Evaluation for Stage 3 of the Construction Sequence 1.9.3.1 In Stage 3 of the construction

sequence, the plant is in a no-mode condition.

The containment dome must be evaluated for both Condition A - With the Full Existence of the Two Construction

Openings, and Condition B - With the Liner Plate Rigged into Place,

Welded, and Leak-Tested.

To ensure that the containment vessel and liner do not fail and adversely impact the spent fuel pool or adjacent safety-related structures during a seismic event, the containment dome must be evaluated as a seismic Category IX/I structure for the following load combination:

LC1S3

= 1.0 D +

1 '

E' 1.0 Fv Since live load is not included in LC1S3, the gravity load or weight due to the temporary platforms and construction equipment will be included in LC1S3 as dead load (D).

1.9.3.2 All the load factors are unity in load combination LC1S3 because, at this stage, the containment serves a

seismic Category II/I function.

For the structural adequacy evaluation of the dome as seismic Category II/I, (a) the allowable capacity of the dome concrete shell will be taken as the nominal

{minimum) ultimate strength per ACI 318-63, Part IV-B; and (b) the allowable capacity of the dome liner system will be taken as 0.9 of its nominal yield strength.

In

addition, the limiting liner (compressive) stresses per Section 1.8.3 will be satisfied.

1.9.3.3 Xn addition to the seismic Category II/I evaluation described

above, the dome structure at this stage

must, also be evaluated for its normal loading condition.

This evaluation will be based on the following two sets of load combinations:

Design Criteria DC-10034A EWR 10034 Page 25 Revision C

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SET 1:

Load Combinations (LC1S3C,

LC2S3C, LC3S3C) for the Dome Concrete Shell LC1S3C

= 1.5 (D + T')

+ 1.8 L + 1.0 Fv LC2S3C

= 1.25 (D + T' L + Sn

+

W)

+ 1.0 Fv LC3S3C

= 0.9 (D + T')

+ 1.1 (Sn

+

W)

+ 1.0 Fv

LC1S3C, LC2S3C, and LC3S3C are based on the required load combinations specified in ACI 318-63, PART IV-B.

Temperature loading, T', will be obtained considering the range of temperatures possible due to factors such as inside and outside temperature, heat transfer coefficient, etc.

SET 2:

Load Combination (LC1S3L) for the Liner System LC1S3L

=

D + L + Sn

+

W + T' Fv 1.9.3.4 The seismic load is not included in any load combination in Sets 1 and 2 in accordance with the State of New York Building Construction Code.

1.9.3.5 The temporary platforms and construction equipment will be included as live load in LC1S3C,

LC2S3C, LC3S3C, and LC1S3L.

1.9.3.6 In this stage of the construction

sequence, the structural evaluation of the concrete shell and liner system of the containment will be performed based on the structural codes and. guidelines specified in Sections 1.8.2 and 1.8.3.

1.9.4 Dome Structure Evaluation for Stage 4 of the Construction Sequence 1.9.4.1 In Stage 4 of the construction

sequence, the plant is returned to a refueling or cold shutdown condition and the containment dome must be structurally evaluated as a seismic Category II/I structure for the following load combination (LC1S4).

LC1S4

= 1.0 D + 1.0 P'

1.0 T

+ 1.0 E' 1.0 Fv

Where, T

and E're obtained the same way as described in Section 1.9.2 for Stage 2.

1.9.4.2 LC1S4 must be considered for both Condition A - Prior to the New Concrete Placement, and Condition B - With the New Concrete in Wet Condition.

The influence of wet concrete in determining the temperature and earthquake loadings on'he liner will be taken into Design Criteria DC-10034A EWR 10034 Page 26 Revision C

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account.

Since live load is not included in LC1S4, the gravity load or weight due to the temporary platforms and construction equipment will be included in LC1S4 as dead load (D)

~

1.9.4.3 In the same manner as Stage 2, for the structural adequacy evaluation of the dome as seismic Category II/I, (a) the allowable capacity of the dome concrete shell will be taken as the nominal (minimum) ultimate strength per ACI 318-63, Part IV-B; and (b) the allowable capacity of the dome liner system will be taken as 0.9 of its nominal ultimate strength.

In

addition, the limiting liner (compressive) stresses per Section 1.8.3 will be satisfied.

1.9 '.4 In addition to the seismic Category II/I evaluation described

above, the dome structure must be evaluated under the normal loading condition.

This evaluation must be performed for both Condition A - Prior to the New Concrete Placement and Condition B - With the New Concrete in Wet Condition.

In the same manner as Stage 3, this evaluation will be performed based on the following two sets of load combinations:

SET 1:

Load Combination (LC1S4C,

LC2S4C, and LC3S4C) for the Dome Concrete Shell LC1S4C

= 1.5 (D + T )

+ 1.8 L + 1.0 Fv LC2S4C

= 1.25 (D + T

+ L + Sn

+ W)

+ 1.0 Fv LC3$ 4C

= 0.9 (D + T')

+ 1. 1 (Sn

+ W)

'+ 1.0 Fv

LC1S3C, LC2S3C, and LC3S3C are based on the required load combinations specified in ACI 318-63, Part IV-B.

SET 2:

Load Combination (LC1S4) for the Liner System LC1S4L =

D + L + Sn

+

W + T

+ Fv 1.9.4.5 The temporary platforms and construction equipment will be included as live load (L) in LC1S4C,

LC2S4C, LC3S4C, and LC1S4L.

The weight of the wet concrete will be included in LC1S4C,

LC2S4C, LC3S4C, and LC1S4L as dead load (D).

1.9.4.6 The seismic load is not included in any load combination in Sets 1 and 2 in accordance with the State of New York Building Construction Code.

1.9.4.7 I

comb Design Criteria Revision C

The structural evaluation of the concrete shell and liner system of the containment dome for all the load inations in Sets 1 and 2 will be performed based on DC-10034A EWR 10034 Page 27 Date 09 26 94

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the structural codes and guidelines specific Sections 1.8.2 and 1.8.3.

1.9.4.8 Regarding the limited SlT, it will be performed after the new concrete attains its design strength or a strength high enough to prevent the damage of the bond between the rebar and concrete to ensure adequate composite action between each individual and each layer of rebar.

As discussed in Section 1.8.4.8, the new concrete mechanical properties are comparable with the existing concrete mechanical properties.

This means the structural evaluation of the containment shell associated with the SIT condition is covered by its original design.

Thus, no additional structural evaluation is required for the SIT condition.

1.9.5 Dome Structure Evaluation for Stage 5 of the Construction Sequence 1.9.5. 1 In Stage 5 of the construction

sequence, the temporary platforms and construction equipment) will be removed from the containment dome.

As discussed in Section 1.8.4, the effects of creep of the new concrete are insignificant, and the new concrete mechanical properties are comparable to the old concrete mechanical properties.

This means the dome is fully restored to its original condition.

Therefore, no additional structural evaluation is required for the following two reasons:

(i) while all or some of the temporary components are still in place, the structural evaluation is covered by the structural evaluation in Stage 1;

and (ii) after the removal of all the temporary components, the. original structural analysis and design of the containment shell remain valid for the restored dome.

1.10.0 1.10.1 Environmental Conditions For the analysis of the containment structure, normal and extreme environmental conditions shall be as specified in the UFSAR.

1.11.0 1.11.1 Interface Re uirements Temporary platforms to be installed on the containment are discussed in Section 2.0.

The temporary removal and reinstallation of interfering portions of the containment spray piping, painter's monorails, and the containment recirculation and cooling system duct are described in Section 6.0.

Design Criteria DC-10034A EWR 10034 Page 28 Revision C

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1.12.0 1.12.1 Material Re uirements Replacement materials for the containment vessel dome shall be compatible with the existing materials.

1.12.2 If a material substitution is required, a

reconciliation back to the original code or standard shall be performed to ensure the new materials meet or exceed the original material requirements.

1.12.3 1.12.3.1 Containment Liner Plate The existing containment liner is %-inch-thick carbon steel plate conforming to ASTM A442-60T Grade 60 with a minimum yield of 32,000 psi.

If needed, replacement liner plate shall be %-inch-thick conforming to ASTM A516 Grade 60 with a minimum yield of 32,000 psi.

1.12.3.2 Welding electrodes for the liner plate shall be E70XX.

Weld end preparation,

welding, and NDE performed in the field shall be in accordance with the Special Processes Manual.'he existing liner plate welds were made from both sides; therefore, no backing bars were used.

For the repair work, welds may be performed from either or both sides provided that, as a minimum, the strength of the weld matches the strength of the base metal.

In the case of a one sided weld, a backing bar shall be used.

1.12.3.3 Existing studs welded to the liner plate and embedded in the concrete are Nelson S6L studs.

Studs and/or an anchorage method of equivalent capacity (e.g.,

stiffeners) shall be used for the re-installation of the existing liner plate or for the installation of new liner plate.

1.12.3.4 Existing test channels for the containment liner are C1Y4 x Y> x

'A fabricated from A36 material.

Replacement test channels shall be channels or bent plate of comparable size fabricated of A36 material.

1.12.3.5 If needed, stiffeners attached to the liner plate sections shall be fabricated from A36 material.

1.12.3.6 The existing paint used on the liner plate is a minimum of 2.5 mil of Carbozinc 511 Gray as manufactured by the Carboline Company.

Painting of replacement liner plate, liner plate attachments, and touch-up painting of the existing liner plate shall be in accordance with RGSE Specification CE-125.

Design Criteria DC-10034A EWR 10034 Page 29 Revision C

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1.12.4 Reinforcing Bar 1.12.4.1

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Existing reinforcing bars are intermediate grade billet steel conforming to A408-62T with a guaranteed minimum yield strength of 40,000 psi.

R placement reinforcing bars shall be ASTM A615 Grade 60.

1.12.4.2 Splicing of the rebar shall be via cadwelds or welded connections.

Cadweld splices shall be in accordance with design output documents and qualification of the splice system shall be in accordance with ASME Section ZZZ, Division 2, Subsection CB, Subparagraph CB-4333.2.

Welding of rebar shall be in accordance with AWS D1.4 and the Special Processes Manual.

1. 12. 5 Welded Wire Fabric The existing welded wire fabric is 6x6-0/0 W.W.F.

conforming to ASTM A185 The replacement material shall be 6x6-0/0 W.W.F.

and conform to ASTM A185.

1.12.6 Miscellaneous Accessories 1.12.6.1 Miscellaneous accessories shall conform to CRSZ MSP-1 for chairs, bolsters, bar support,

spacers, and tie wire.

1.12.7 1.12.7.1 Containment Concrete The existing containment concrete was required to achieve a compressive strength of 5000 psi in 28 days.

The replacement concrete shall achieve a compressive strength of 5000 psi at 7 days.

The replacement concrete shall meet the requirements of ASME/ANSZ N45.2.5.

1.13.0 1.13.1 1.14.0 1.14.1 Mechanica1 Re uirements Not Applicable.

Structural Re uirements The modified containment dome shall be structurally adequate to carry the defined loads during the steam generator replacement outage.

The restored containment dome shall meet or exceed the original structural capacity of the undisturbed containment dome.

Restoration of the containment dome shall not modify the load carrying characteristics of the structure.

1. 14. 3 US 3.

Design Criteria Revision C

Weld seams in the liner plate are covered with a test channel to permit testing of leaktightness.

Test channels shall be welded where plate sections meet

'ng welds that meet the original structural capacity.

DC-10034A EWR 10034 Page 30

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1.15.0

'o 1.15.1

~

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1.16.0 1.16.1

1. 17. 0

1. 17. 1 1.18.0 1.18.1 H draulic Re uirements Not Applicable.

Chemistr Re uirements Not Applicable.

Electrical Re uirements Not Applicable.

La out and Arran ement Re uirements The dome openings shall be of sufficient size to allow the egress/entry of the steam generators from/into the containment dome.

1.18.2 An analysis shall be performed to establish the size and location of the dome openings to ensure that the postulated drop of a steam generator during rigging into or out of the containment does not result in impact on the spent fuel pool.

1.19.0 1.19.1 1.20.0 0 erational Re irements Not Applicable.

Instrumentation and Control Re irements 1.20.1 1.21.0 1.21.1 1.22.0 1.22.1 1.23.0 1.23 '

Not Applicable.

Access and Administrative Control Re uirements Not Applicable.

Redundanc Diversit and Se aration Re irements Not Applicable.

Failure Effects Re irements An analysis shall be performed to establish the size and location of the dome openings to ensure that the postulated drop of a steam generator during rigging into or out of the containment does not result in impact on the spent fuel pool.

1.24.0 1.24.1 1.24.1.1 Test Re uirements Liner Plate Prior to the placement of concrete, NDE and leak testing of the liner plate shall be conducted in Design Criteria DC-10034A EWR 10034 Page 31 Revision C

Date 09 26 94

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1. 24. 2 accordance with the Special Processes Manual and design output documents and acceptable, test results achieved.

Containment Concrete

1. 24. 2. 1 Concrete and constituent testing shall be performed in accordance with ASME/ANSI N45.2.5.

1.24.2.2 To confirm the structural integrity of the containment, a limited, post-closure, structural integrity test (SIT) shall be conducted to a pressure of 69 psig (1154 of the containment design pressure of 60 psig) in accordance with UFSAR Section 1.8.1.1'8.1 and NRC Safety Guide 18.

This limited SIT will be limited, to the area of the construction openings.

The replaced portions of the containment dome shall be appropriately instrumented with strain gages;

however, instrumentation for the measurement of the global deformation of the containment need not be provided.

1.24.3 Reinforcing Bar 1.24.3.1 Testing of the replacement reinforcing bar in the containment dome shall be performed in accordance with ASME/ANSI N45.2.5.

1.24.3.2

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Cadweld splice and splicer testing shall be performed in accordance with ASME Specification SA-370,.

Per UFSAR Section 3.8.1.6.3, the cadweld splices shall develop. the minimum ultimate (tensile) strength of the higher-strength, replacement rebar.

1.24.3.3 Testing of rebar splices via welded connections shall be in accordance with AWS D1.4 and the Special Processes Manual.

1.25.0 Accessibilit Maintenance Re air Re lacement Pum and Valve Test Pro ram and Inservice Ins ection Re uirements 1.25.1 The replacement liner plate shall have a test channel welded where plate sections meet using welds that meet the original structural capacity.

1.25.2 Repair and replacement activities shall be in accordance with the Bechtel Quality Assurance Program Plan.

Personnel Re uirements 1.26.0 1.26.1 and Design Criteria Revision C

Personnel performing welding and nondestructive examination shall be qualified in accordance with the AWS Code and the ASME Code, Sections III, V, IX, XI, SNT-TC-1A, as applicable.

DC-10034A EWR 10034 Page 32 Date 09 26 94

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1.27.0 Trans ortabilit Re uirements 1.27.1

~

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1.28.0

1. 28. 1 Not Applicable.

Fire Protection Re uirements The modifications planned to be made to the containment

concrete, reinforcing bar, and liner plate are one-for-one replacements.

Therefore, there will be no permanent increase in the probability or consequences of a fire and there will be no permanent increase in combustible loadings.

1.28.2 No existing fire barriers or fire barrier penetration seals will be affected and no new fire barri.ers will be installed.

1.28.3 Non-combustible and heat resistant materials will be used whenever possible and to the extent practical.

1.28.4 The planned modifications to the containment dome shall not adversely affect the performance of any existing fire protection equipment.

1.28.5 Temporary fire protection provisions during the installation,

use, and removal of the temporary structures shall be established.

1.28.6

~

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An Appendix R Conformance'erification shall be performed to ensure that. this modification does not adversely impact existing Appendix R Compliance methods.

1.29.0 1.29.1 Handlin Re uirements Handling requirements for the liner plate sections are identified in Design Criteria Document DC-10034B.

1.30.0 1.30.1 Public Safet Re irements Modifications to the dome shall not adversely affect offsite doses associated with any postulated accident or normal outage levels.

1. 30. 2 A means of temporarily closing the dome openings (e.g.,

tarpaulin covers) shall be provided and available.

1.31.0 1.31.1 A licabilit All weld end preparation,

welding, and NDE performed in the field shall be in accordance with the Special Processes Manual.

1. 31. 2

~

~

Material requirements are identified in Section 1.12.0.

Design Criteria DC"10034A EWR 10034 Page 33 Revision C

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1.31.3 1.32.0 1.32.1 Concrete and rebar shall be removed by chipping, rock drilling, or saw cutting.

The method of removal shall not adversely affect the structural adequacy of the remainder of the containment dome that will not be removed and replaced.

Personnel Safet Re uirements The design of the modifications to the dome shall include precautions to ensure that any occupational radiation exposures to personnel are As Low As Reasonably Achievable (ALARA).

1.32.2 Nork shall be performed in accordance with OSHA requirements.

1.32.3 1.33.0 1.33.1 A Pre-ALARA Review Checklist will be completed prior to implementation of this work.

Environmental uglification Re uirements A 10CFR50.49 (EQ) Applicability checklist has been completed and is included as Attachment 1.

1.34.0 Uni e

Re irements 1.34.1

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~

A modal analysis of the containment dome will be performed to investigate the range of its vibration frequencies based on possible variations of concrete stiffness (cracked versus uncracked properties).

The resulting mode shapes will be studied to determine if any are excitable due to the operating frequencies of the automated hydraulic jackhammers.

Design Criteria DC-10034A EWR 10034 Page 34 Revision C

Date 09 26 94

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4

SECTION 2.0 TEMPORARY PLATFORMS 2.1.1 2.1.1.1 Summa Descri tion of the Desi n Summary Temporary facilities planned to be erected in the dome area to facilitate work on the containment include:

~

Various structural steel platforms or frames will be installed on the containment dome to support the automated hydraulic jackhammers, temporary service cranes, craft personnel, and miscellaneous construction equipment and materials.

~

Temporary equipment and material laydown areas will be erected on the facade structure and the intermediate building roof.

~

A temporary craft break area, stairwell(s),

and personnel walkways will be erected on the facade structure and intermediate building roof.

2. 1.1.2 These temporary structures will be erected prior to plant shutdown and will not be removed until after the plant has returned to operation following replacement of the steam generators.

2.1. 1.3 This section of the Design Criteria defines the requirements for the design and analysis of the temporary support structures

before, during, and after the steam generator replacement (SGR) outage.

2. 1. 1.4 Temporary electrical power for these facilities is addressed in Design Criteria Document DC-10034D.

2.1.2

2. 1.2. 1=

Functions The function. of the temporary dome'latforms is to facilitate the construction activities planned to be performed and to provide a safe means of worker access to the dome openings.

2.1.2.2 The function of the temporary facade laydown platform is to provide a laydown area close to the dome openings for equipment and materials.

2. 1.2.3 The function of the temporary craft break area is to provide a break area and lunch room for personnel working on the containment dome.

Design Criteria DC-10034A EWR 10034 Page 35 Revision C

Date 09 26 94

5 C,

2.1.3 2.1.3.1 Performance Requirements Performance requirements for the work platforms are to provide a safe and adequate means of access to the dome openings and to provide adequate space and support of construction activities in the dome area.

2.1.3.2 Performance requirements of the temporary facade laydown structure are to provide support for equipment

~ and material loads near the dome area during the SGR outage.

2.1.3.3 Performance requirements of the temporary craft break area are to provide adequate break area and lunch room facilities for personnel working on the containment dome.

2.1.4 2.1.4.1 2.1.5 2.1.5.1 Control Not Applicable.

Modes of Operation The temporary work platforms, laydown area, and craft break area shall be capable of performing their functions under all modes of plant operation.

2.2.0

~

~

Referenced Documents Bechtel Quality Assurance Program Plan for Rochester Gas 6 Electric Corporation, R.

E. Ginna Nuclear Power

Plant, Steam Generator Replacement

Project, Bechtel Job No.

22225.

2.2.2 Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

2.2.3

.2.2.3.1 2.2.4 2.2.4.1 2.2.5 2.2.5.1 American Concrete Institute (ACI) 318-63, Building Code Requirements for Reinforced Concrete and Commentary.

American institute of Steel Construction (AISC)

Manual of Steel Construction, 9th edition.

American Welding Society (AWS)

D1.1-94, Structural Welding Code.

2.2.6 Special Processes

Manual, Ginna Nuclear Power Plant Steam Generator Replacement

Project, Bechte1 Job No.

22225.

Design Criteria DC-10034A EWR 10034 Page 36 Revision C

Date 09 26 94

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jl I

'C,~J s 4

,s'

2.2.7 2.2.7.1

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~

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2.2.7.2 American Society for Testing and Materials (ASTM)

ASTM A36, Revision A-93, Standard Specification for Structural Steel ASTM A307, Revision A-93, Standard Specification for Carbon Steel Bolts and Studs 60,000 psi Tensile Strength.

2.2.7.3 2.2.7.4 2.2.7.5 2.2.7.6 2.2.7.7 2.2.7.8

~

~

~

2.2.8 2.2.8.1, ASTM A325-93, Standard Specification for Structural Bolts, Steel, Heat Treated 120/105 ksi Minimum Tensile Strength.

ASTM A446-,93, Standard Specification for Sheet Steel, Zinc-Coated (Galv) By.the Hot Dip Process, Structural (Physical Quality).

ASTM A490-93, Standard Specification for Heat Treated Steel Structural Bolts, 150 ksi Tensile Strength.

ASTM A563-93-93, Standard Specification for Carbon and Alloy Steel Nuts.

.ASTM A569, Revision A-91, Specification for Steel, Carbon

(.15 Max Percent)

Hot-Rolled Sheet and Strip, Commercial Quality.

ASTM A659-92, Specification for Steel, Carbon

(.16 Min to.25 Max Percent)

Hot-Rolled Sheet and Strip, Commercial Quality.

American Society of Civil Engineers (ASCE)

'ASCE 7-93, Minimum Design Loads for Buildings and Other Structures.

2.2.9 2.2.9.1 2.2.10 Bechtel Design Guides EDG-C0313, Revision 1, Civil/Structural Engineering Design Guide for Concrete Expansion Anchors.

RGEE Specification CE-125, Technical Requirements for Furnishing and Erecting Structural Steel, Revision 7.

Seismic Upgrade

Program, EWR 2512, Revision 5, dated 4/11/89.

E Seismic Cate or 2.2.11 2.3.0 2.3.1 aaf Design Criteria Revision C

The temporary platforms or frames will be attached to the containment

dome, intermediate building roof, or the facade structure.

The containment

dome, intermediate building, and facade are classified as ety-related, seismic Category I structures.

DC-10034A EWR 10034 Page 37 Date 09 26 94

2'.2 To ensure the temporary platforms do not fail and impact adjacent safety-related equipment during a seismic event, they will be designed and classified as safety significant, seismic Category II/I structures.

Miscellaneous construction equipment in the dome area shall also be supported and/or restrained, as needed, to preclude impact on adjacent safety-related equipment during a seismic event.

2.3.3 Anchor bolts installed in the containment dome or intermediate building concrete are classified as safety significant, seismic Category II/I.

The process of installing and removing anchor bolts is classified as safety-related to ensure no adverse effects on the concrete.

Grout used.to repair seismic Category I concrete surfaces is classified as safety-related.

Cutting of rebar is prohibited unless authorized by Engineering and the structural adequacy of the affected structure with cut rebar has been confirmed by an engineering evaluation and/or analysis..

2.3 '

All connections and modifications to the facade structure are classified as safety-related.

2.3.5 All coatings for the outside temporary structures are classified as nonsafety-related.

2.4.0 2.4.1 2.5.0 2.5.1 2.6.0 2.6.1 ualit Grou Code Class Fluid S stems Onl Not Applicable.

Electrical S stem Safet Classifications Not Applicable.

A Pro ram A licabilit The design, installation,

use, and removal of the temporary structures shall be performed in accordance with the Bechtel Quality Assurance Program Plan.

2.7.0 Codes Standards and Re

lator, Re uirements The following principal codes, standards, and regulatory requirements shall be used in the design of the temporary structures:

2.7. 1 Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

2.7.2 2.7.2.1 American Concrete Institute (ACI) 318-63, Building Code Requirements for Reinforced Concrete and Commentary.

h Design Criteria DC-10034A EWR 10034 Page 38 Revision C

Date 09 26 94

2.7.3 2.7.3.1

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2.7.4 2.7.4.1 2.7.5 2.7.5.1 American Institute of Steel Construction (AXS Manual of Steel Construction, 9th edition.

American Welding Society (AWS)

D1.1-94, Structural Welding Code.

American Society of Civil Engineers (ASCE)

ASCE 7-93, Minimum Design Loads for Buildings and Other Structures.

2.8.0 2.8.1 Desi n Conditions

~ The -temporary structures will be erected before the SGR outage begins and will not be removed until after return to power operation.

Thus, the temporary structures shall be considered as live load for the containment

dome, intermediate building, and facade.

These supporting structures shall be evaluated using their original design codes.

2.8.2 The temporary platforms shall be consistent with the Design Criteria for EWR 2512.

2.8

~ 3

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The design and use of the temporary structures shall consider the effects of tornado winds and resulting tornado missiles in accordance with the UFSAR.

2.9.0 2.9.1 2.9.1.1 Load Conditions Load. Combinations The following combinations of loading conditions shall be investigated and the most critical loading combinations used in the design of structural steel platforms mounted to the containment,

facade, and intermediate building roof:

(1)

Dead Load

+ live load Q =

D + L (2)

Dead load

+ live load

+ wind load Q

=

D + L +

W (3)

Dead load

+ live load

+ earthquake Q=D+L+E'esign Criteria DC-10034A EWR 10034 Page 39 Revision C

Date 09 26 94

l II 5

.i+i fp T

WHERE:

D =

dead load of the structure E'

safe shutdown earthquake load 1

L =

live loads

~

minimum live load of 200 psf

~

equipment load

~

snow load per ASCE 7-93 W =

wind loads per ASCE 7-93 Q

=

load combination to be used with allowable stress design The effect of tornado winds and tornado missile impacts shall be considered.

2.9.1.2 Allowable Stresses (1)

Dead Load

+ live load Allowable stresses shall be as defined in the AISC Manual of Steel Construction, 9th edition.

(2)

Dead load +'live load

+ wind load Allowable stresses shall be as defined in the AISC Manual of Steel Construction, 9th edition, with an increase factor of 1.33.

(3)

Dead load

+ live load

+ earthquake Allowable stresses shall be as defined in the AISC Manual of Steel Construction, 9th edition, with an increase factor of 1.6 but'hall not exceed 0.9 of the yield stress.

2.9.2 Anchor Bolt Design The design of new concrete anchorages to hardened concrete shall use Hilti Kwik Bolt II (HKB II) or Drillco Maxi-Bolt (DMB) anchors unless a special application requires the use of another type of anchor.

Anchor bolts shall be designed in accordance with Bechtel Design Guide EDG-C0313.

In addition, the criteria provided in Appendix A to this document shall be considered for any safety-related anchorage applications.

Design Criteria DC-10034A EWR 10034 Page 40 Revision C

Date 09 26 94

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2.10.0

2. 10. 1

2. 10. 2 2.11.0 2.11.1 2.11.2 Environmental Conditions The temporary structures shall be designed to withstand all anticipated environmental conditions including

wind, snow, and temperature.

For the design and analysis of the temporary platforms, normal environmental conditions shall be in accordance with ASCE 7-93.

Tornado winds and tornado missile impacts shall be as specified in the UFSAR.

Interface Re uirements The temporary platforms and structures will be attached to the containment

dome, intermediate building roof, or the facade structure and will be in place before,

during, and after the SGR outage.

The containment

dome, intermediate building roof, and facade shall be checked for the addition of the temporary structures as a live load.

This evaluation shall be performed in accordance with the existing design codes.

2.12.0

2. 12. 1

2. 12.2 2.12.3 2.12.4 Material Re irements Structural steel Bolts Nuts Anchor bolts ASTM A36 ASTM A307, ASTM A325, or ASTM A490 ASTM A563 Hilti Kwik Bolt II or Drillco Maxibolts 2.12.5 Sheeting/decking ASTM A446, A569 or ASTM A659 2.12.6 2.12.7 2.12.8 Non-skid grating Checkered plate Welding Materials ASTM A556 or ASTM A659 ASTM A36 or ASTM A569 E70XX electrodes.

All welding performed in the field shall be in accordance with the Special Processes Manual.

2.12.9

2. 12. 10 Grout Coatings Masterflow 713 or 928 by Master Builders Company In accordance with RGsE Specification CE-125 Design Criteria DC-10034A EWR 10034 Page 41 Revision C

Date 09 26 94

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2. 13. 0

2. 13. 1 Mechanical Re uirements Not Applicable.

2.14.0 2.14.1 Structural Re uirements The temporary structures shall be designed to the AZSC Manual of Steel Construction, 9th edition.

2. 14. 2 The temporary structures shall be capable of supporting the required loads without jeopardizing the integrity of the containment, intermediate building roof, or the facade.

2.15.0 2.15.1 2.16.0 2.16 '

2. 17.0

2. 17. 1 H draulic Re uirements Not Applicable.

Chemist Re uirements Not Applicable.

Electrical Re irements Temporary electrical power for these structures is discussed in Design Criteria Document DC-10034D.

2.18.0

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La out and Arran ement Re irements 2.18.1 The location and layout of the temporary structures shall provide a safe means of access to the dome openings.

The layout and arrangement of the platforms shall facilitate the construction activities planned to be performed and provide a safe and efficient means of access between the individual work areas.

2.19.0'.19 0 erational Re irements The temporary platforms and laydown area may be installed prior to shutdown for the steam generator replacement (SGR) outage.

These structures will be operational as needed to support SGR activities.

These structures may be removed from the containment after returning to power following the SGR.

2.20.0 2.20.1 2.21.0

2. 21. 1 Instrumentation and Control Re irements Not Applicable.

Access and Administrative Control Re irements Personnel access to the dome area will be via the stairwell located in the northwest corner of the Design Criteria DC-10034A EWR 10034 Page 42 Revision C

Date 09 26 94

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Intermediate Building between column lines 3 and 3a and F and F6.

2.22.0 Redundanc Diversit and Se aration Re uirements 2.22.1 2.23.0 2.23.1 Not Applicable.

'Failure Effects Re uirements The construction,

use, and removal of the temporary structures in the containment, dome area shall not impair normal plant operation, the continued operation of the plant following an Operating Basis Event, or the ability to safely shut down the plant following a Design Basis Earthquake.

2.24.0 2.24.1 2.25.0 2.25.1 2.26.0 2.26.1 2.27.0 2.27.1 2.28.0 2.28. 1 Test Re uirements Not Applicable.

Accessibilit Maintenance Re air Re lacement Pum and Valve Test Pro ram and Xnservice Ins ection Re uirements Not Applicable.

Personnel Re uirements Not Applicable.

Trans ortabilit Re uirements Not Applicable.

Fire Protection Re uirements The platforms, laydown area, craft break area, etc.

are temporary structures; therefore, there will be no permanent increase in the probability or consequences of a fire and there will be no permanent increase in combustible loadings.

2.28.2 No existing fire barriers or fire barrier penetration seals will be affected and no new fire barriers will be installed.

2.28.3

.The design and fabrication of the temporary structures shall use non-combustible and heat resistant materials whenever possible and to the extent'ractical.

2. 28. 4

.The design and use of the temporary structures shall not adversely affect the performance of any existing fire protection equipment.

Design Criteria DC-10034A EWR 10034 Page 43 Revision C

Date 09 26 94

2.28.5 Temporary 'fire protection provisions during t e installation,

use, and removal of the tempora structures shall be established.

Welding and grinding permits shall be obtained prior to work and any special fire protection measures, taken as required by the permits.

2.'8. 6 An Appendix R Conformance Verification shall be performed to ensure that this modification does not adversely impact existing Appendix R Compliance methods.

2.29.0 2.29.1 Handlin Re uirements Handling.requirements. for the materials used to construct the temporary, structures are identified Design Criteria Document DC-10034B.

2.29.2 Appropriate pre-assembly of the temporary structures shall be performed to facilitate rapid installation in the dome area.

2.30.0 2.30.1 2.31.0 Public Safet Re uirements Not Applicable.

A licabilit

2. 31. 1

~

~

All weld end preparation,

welding, and NDE performed in the field shall be in accordance with the Special Processes Manual.

2.31.2 Material requirements for the temporary structures are identified in Section 2.12.0.

2.32.0 2.32.1 Personnel Safet Re uirements The design, installation,

use, and removal of the temporary structures shall minimize the potential for personnel injury. All work shall be performed in accordance with OSHA requirements.

2.32.2 A Pre-ALARA Review Checklist will be completed prior to implementation of this work.

2.33.0 2.33.1 Environmental uglification Re irements A 10CFR50.49 (EQ) Applicability checklist has been completed and is included as Attachment 1.

2.34.0 2.34.1 Uni ue Re irements Not Applicable.

Design Criteria DC-10034A EWR 10034 Page 44 Revision C

Date 09 26 94

E

SECTION 3.0 PERMANENT STEAM GENERATOR SUPPORTS 3.1.0

~

~

Summar Descri tion of the Desi n 3.1.1 3.1. 1. 1 Summary Description of Supports A description of the steam generator, reactor coolant

pump, and RCS pipe supports is as follows:

The steam generator upper lateral support configuration consists of a steel ring attached to the compartment walls by 1 snubber set and 3 or 4

sets of struts.

To account for thermal movements, the steam generators are cold-jacked in the Cold Shutdown condition to gaps set appropriately for the hot position.

The steam generator lower support configuration consists of four columns pinned top and bottom, with a support bracket assembly that pins to the support pad of the steam generator.

Shims are placed in-between the support bracket assembly and the clevis assembly on top of the lower support column.

The lower lateral steam generator

support, system is provided for seismic restraint and consists of steel beams attached to the compartment walls with bumper pockets that are shimmed to each steam generator support pad.

~

The reactor coolant pump support system is similar to the steam generator support system with three columns pinned top and bottom.

Lateral support is provided by two tie rods that are bolted to a pump foot and extend to the shield wall concrete.

~

The RCS pipes have no restraints but are supported on the steam generator, reactor coolant

pump, and reactor vessel nozzles.

Crossover leg bumpers (whip restraints) are installed.

3.1.1.2 3.1.1.2.1 Steam Generator Removal/Reinstallation Method Steam Generator Removal Sequence The following is a description of the planned sequence for removal of each old steam generator:

~

Release the cold-jacked condition at the upper lateral support (ring).

Design Criteria DC-10034A EWR 10034 Page 45 Revision C

Date 09 26 94

gA A ~

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Remove struts and snubbers and install adjustable struts on the upper lateral support (ring).

~

Temporarily restrain the steam generator lower support columns.

Shim the steam generator support pads in the lower lateral support bumper pockets to preclude steam generator movement after RCS severance.

Shim the primary coolant crossover leg piping at the existing bumper and crib under the crossover leg piping to provide deadweight support.

Deadweight support of the hot leg piping is not anticipated.

Temporarily support other steam generator-connected piping and tubing as required.

~

Sever the primary coolant, main steam, feedwater, and other piping/instrumentation connected to the steam generator.

~

Assess the extent of RCS pipe movement against machining and welding tolerances, and, if unacceptable, make decision to replace the cold and/or hot leg elbows.

Remove the temporary adjustable struts on the upper lateral support (ring).

~

Connect the Transi-Lift to the steam generator.

Loosen support bracket bolts.

Ensure that the Transi-Lift is carrying the load prior to cutting brackets.

~

Cut off the support brackets just below the lower lateral support structure.

~

Raise the steam generator to clear the lower lateral support structure.

Xnstall primary coolant system nozzle covers.

~

Rig out the steam generator.

Design requirements for steam generator-connected piping and tubing are identified in Design Criteria Document DC-10034C

~

Steam generator rigging requirements are identified in Design Criteria Document DC-10034B.

3.1.1.2.2 Steam Generator Installation Sequence The following is a description of the planned sequence for installation of each new steam generator:

Design Criteria DC-10034A EWR 10034 Page 46 Revision C

Date 09 26 94

I

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Install the new upper bracket assembly steam generator support foot pad by installing the support pad pin, keeper plates, and bolts.

Install the new lower bracket assembly on the lower support column by loosely installing the four bolts.

Rig in the new steam generator.

Adjust the orientation of the steam generator nozzles as necessary to achieve acceptable fit-up to the primary coolant system elbows, Elevation changes will be accomplished by adjusting the thickness of shim stacks between the new lower bracket assembly and the steam generator support column and/or between the upper and lower bracket assemblies.

Lateral adjustments will be accomplished using wedges in the lower lateral support structure and come-a-longs.

Load the support columns with the weight of the steam generator.

Install and tighten the bolts between the upper and lower bracket assemblies.

Tighten the bolts between the lower bracket assembly and the support column.

Connect the.temporary adjustable struts on the upper lateral support (ring).,

Shim the steam generator support pads in the lower lateral support bumper pockets to fix the location of the bottom of the steam generator.

Reconfirm primary coolant system nozzle-to-elbow fit-up.

Disconnect all rigging from the steam generator.

Initiate welding of the primary coolant system.

After 1/3 weld-out is completed, remove all temporary shims and restraints in the upper and lower lateral supports.

Disconnect and remove the temporary support column restraints.

Once all welding of steam generator-connected piping and instrumentatation tubing is completed, remove piping temporary supports.

Cold-jack the steam generator at the upper lateral support (ring) to the specified position.

Design Criteria DC-10034A EWR 10034 Page 47 Revision C

Date 09 26 94

~

Connect the upper lateral support (ring) snubbers and struts..

Design requirements for steam generator-connected piping and tubing are identified in Design Criteria Document DC-10034C.

Steam generator rigging requirements are identified in Design Criteria Document DC-10034B.

3.1.1.3 3.1.1.3.1 New Steam Generator Upper Supports (Rings)

Xn order to facilitate the installation of the new steam generators, new upper supports (restraint rings) will be fabricated and installed during the steam generator replacement.(SGR) outage.

The new upper supports will be a like-for-like replacement with no change in the basic support configuration.

3. 1. 1. 4 3.1.1.4.1 Snubber Hydraulic Tubing and Reservoirs As discussed in Section 7.0 of this document, portions of the existing secondary manway access platforms will be temporarily removed and re-installed during the steam generator replacement outage.

The hydraulic fluid reservoirs and portions of the hydraulic tubing serving the upper lateral support (ring) snubbers are supported from the existing, secondary manway access platforms and,

thus, must also be temporarily removed and re-installed.

3. 1.1.5 New Steam Generator Lower Support Bracket Assemblies s

Shims

3. 1. 1.5. 1 To facilitate the removal and reinstallation of the steam generators, new support bracket assemblies which connect the steam generator support foot pads to their respective vertical support columns will be designed, fabricated, and installed.

3. 1. 1. 5. 2 The existing, one-piece support bracket assembly will be replaced with two pieces:

an upper bracket assembly and a lower bracket assembly.

This arrangement will allow the upper bracket assembly to be bolted and pinned to the steam generator support foot pad and the lower bracket assembly to be bolted to the support column clevis assembly prior to the steam generator being installed in the cubicle.

The upper and. lower support bracket assemblies will subsequently be bolted together.

3.1.1.5.3 New pins will be fabricated to connect the upper support bracket assembly to the steam generator support foot pad.

Design Criteria DC-10034A EWR 10034 Page 48 Revision C

Date 09 26 94

3. 1. 1. 5.4 New bolts,

washers, and nuts will be procure connect the lower support bracket assembly to the lower support column and to connect the upper and lower bracket assemblies together.

3.1.1.5.5 Shims of varying thicknesses will be procured to adjust the height of the steam generator nozzles so as to achieve fit-up to the reactor coolant piping.

These shims will be placed between the upper and lower support bracket assemblies and/or between the lower bracket assembly and the clevis assembly on top of the support column.

3.1.1.5.6 3.1.1.6 New keeper plates and brackets will be fabricated.

This section of the Design Criteria defines the requirements for the design,

analysis, fabrication, and installation of the steam generator permanent supports.

3. 1. 1.7 Requirements for the design and analysis of the temporary restraints of the steam generator support columns are identified in Section 4.0 of this Design Criteria Document.

3. 1. 1.8 Requirements for the design and analysis of. temporary supports and/or restraints of the RCS piping are identified in Design Criteria Document DC-10034C.

Functions 3.1.2.1 The function of the upper support (ring) is to prevent excessive movement of the steam generator in any horizontal plane.

3. 1.2.2 The function of the snubber hydraulic tubing and reservoirs is to provide a supply of hydraulic fluid to the upper support (ring) snubbers.

3;1.2.3 The function of the steam generator lower supports is to provide deadweight support and seismic restraint (download and overturning restraint) of the steam generators during all modes of reactor operation.

The steam generator support columns are also designed to permit free thermal growth.

3. 1.3

3. 1.3.1 Performance Requirements The steam generator upper support (ring), snubber hydraulics, and lower supports must satisfy their performance requirements as identified in the UFSAR.

Design Criteria DC-10034A EWR 10034 Page 49 Revision C

Date 09 26 94

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3. 1.4 3.1.4.1 Control The removal and installation of the permanent steam generator supports will not interface with any existing control systems.

Therefore, there is no requirement for additional controls on any existing systems,

features, or equipment.

3.1.5 3.1.5.1 Modes of Operation The existing steam generator upper support (ring) and lower supports must remain functional in all modes of reactor operation and until the reactor vessel has been defueled and the reactor cavity drained.

3.1.5.2 The new steam generator upper support (ring) and lower supports shall be installed and functional prior to refueling.

3.1.5.3 No work shall be performed on any system or component until it has been released for work by Operations.

3.2.0 3.2.1 Referenced Documents Bechtel.Quality Assurance Program Plan for Rochester Gas 6 Electric Corporation, R.

E. Ginna Nuclear Power

, Plant, Steam Generator Replacement

Project, Bechtel Job No.

22225.

3.2.2

~

~

Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

3.2.3 Special Processes

Manual, Ginna Nuclear Power Plant Steam Generator Replacement

Project, Bechtel Job No.

22225.

3.2.4 3.2.4.1 3.2.4.2 3.2.5 3.2.5.1 3.2.6 American Institute of Steel Construction (AISC)

Manual of Steel Construction, 8th edition.

Manual of Steel Construction, 9th edition.

American Welding Society (AWS)

D1.1-94, Structural Welding Code.

ASME Boiler a Pressure Vessel

Code,Section III, 1974 Edition, Subsection NF and Appendix F.

3.2.7 "Evaluation of the Reactor Coolant System for the Steam Generator Hydraulic Snubber Replacement Program,"

Westinghouse Electric Corporation, September 1992.

Design Criteria DC-10034A EWR 10034 Page 50 Revision C

3.2.8 3.2.9

~

~

3.2.9. 1 3.2.9.2

3. 2.9. 3 3.2.10 3.2.10.1 3.2.10.2 3.2.10.3

3. 2. 10. 4 3.2.10.4 3.2.10.5 3.2.10.6 3.2.11 3.2.12 3.2.13 3.3.0 3.3.1 Design Analysis, Snubber Replacement Program Review RCS Embedments, EWR 1483, Revision 1.

Station Drawings 4155-D-521-049, Steam Generator and Reactor Coolant Pump Lower Supports.

4155-D-521-050, Steam Generator and Reactor Coolant Pump Intermediate and Upper Supports.

EWR 1483-33013-1979, Steam Generator Upper Supports Rigid Struts

- Shop Details Sheets 1

4.

American Society for Testing and Materials (ASTM)

ASTM A36, Revision A-93, Standard Specification for Structural Steel ASTM A194, Revision A-93, Standard Specification for Carbon and Alloy Steel Nuts for Bolts for High-Pressure and High-Temperature Service.

3 ASTM A490-93, Standard Specification for Heat Treated Steel Structural Bolts, 150 ksi Tensile Strength.

ASTM A514-93 (USS T-1), Standard Specification for High-Yield-Strength,

Quenched, and Tempered Alloy Steel

Plate, Suitable for Welding.

ASTM A516-90, Standard Specification for Pressure Vessel Plates, Carbon Steel, for Moderate-and Lower-Temperature Service.

ASTM A563-93-93, Standard Specification for Carbon and Alloy Steel Nuts.

ASTM F436-93, Specification for Hardened Steel Washers.

RGSE Specification CE-125, Technical Requirements for Furnishing and Erecting Structural Steel, Revision 7.

RG&E Specification ME-245, Installation of Class 2 and 3 Nuclear Safety-Related Tubing Systems, Rev.

2 dated February 4,

1993.

Seismic Upgrade

Program, EWR 2512, Revision 5, dated 4/11/89.

Seismic Cate or The steam generator upper and lower supports and all replacement components are classified as safety-

related, seismic Category I structures.

Design Criteria DC-10034A EWR 10034 Page 51 Revision C

Date 09 26 94

Wf

3.3.2 3.3.3 3.4.0 3.F 1 The steam generator snubber hydraulic tubing, reservoirs,

supports, and associated components are classified as safety-related, seismic Category I.

Coatings are classified as safety-related.

ualit Grou Code Class Fluid S stems Onl The existing steam generator supports were designed and fabricated to the requirements of the AISC Manual of Steel Construction, 8th edition.

Per UFSAR Section 3.9.3.2.2 and Reference 3.2.7, the supports were subsequently re-evaluated in accordance with the ASME B&PV Code, 1974 Edition,Section III, Subsection NF and Appendix F

~

3.5.0 3.5.1 3.6.0 3.6.1 Electrical S stem Safet Classifications Not Applicable.

A Pro ram A licabilit All work associated

design, fabrication, and installation of the permanent steam generator supports shall be performed in accordance with the Bechtel Quality Assurance Program Plan.

3.7.0

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Codes Standards and Re lator Re uirements The following principal codes, standards, and regulatory requirements shall be used in the design of the permanent steam generator support modifications:

3

~ 7.1 Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

3.7. 1 3.7.1.1 3.7.2 3.7.2.1 3.7.3 American Institute of Steel Construction (AISC)

Manual of Steel Construction, 8th edition American Welding Society (AWS)

D1.1-94, Structural Welding Code.

ASME Boiler 6 Pressure Vessel

Code,Section III, 1974 Edition, Subsection NF and Appendix F.

3.7.4 "Evaluation of the Reactor Coolant System for the Steam Generator Hydraulic Snubber Replacement Program,"

Westinghouse Electric Corporation, September 1992.

Design Criteria DC-10034A EWR 10034 Page 52 Revision C

Date 09 26 94

3.8.0 3.8.1 Desi n Conditions Consistent with the existing steam generator

supports, the replacement steam generator support components shall be fabricated to the requirements of the AISC Manual of Steel Construction, 8th edition.

3.8.2 Consistent with UFSAR Section 3.9.3.2.2 and Reference 3.2.7, the replacement steam generator supports shall be evaluated in accordance with the ASME BGPV Code, 1974 Edition,Section III, Subsection NF and Appendix F.

3.9.0 3.9.1 Load Conditions Consistent with UFSAR Section 3.9.3.2.2 and Reference 3.2.7, the new steam generator support components shall be evaluated in accordance with the ASME BGPV Code, 1974 Edition,Section III, Subsection NF and Appendix F.

3.10.0 3.10.1 Environmental Conditions Environmental conditions for the design and analysis of the permanent steam generator support components shall be in accordance with the UFSAR.

3.11.0 3.11.1 Interface Re irements The steam generator upper supports interface with the steam generator shell and the snubbers/struts attached to the steam generator compartment walls.

3. 11.2 The steam generator support bracket assemblies interface with the steam generator support foot pads, the lower lateral supports, and the lower support columns.

3.12.0 3.12.1 Material Re irements Replacement materials for the steam generator supports and snubber hydraulic components shall be.compatible with the existing materials.

3.12.2 If a material substitution is required, a

reconciliation back to the original code or standard shall be performed to ensure the new materials meet or exceed the original material requirements.

3.12.3 3.12.4 Upper Support Upper Support Pins ASTM A36 or ASTM A516 Grade 70 ASTM A193 Grade B7 DC-10034A EWR 10034 Page 53 Revision C

Date 09 26 94

3. 12. 5 3.12.5.1

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Lower Supports Support Bracket Assemblies USS T-1 (ASTM A514 Grade F) 3.12.5.2 3.12.5.3 3.12.5.4 3.12.5.5 3.12.5.6 3.12.6 Pins Bolts Nuts Washers Shims Coatings USS T-1 (ASTM A514 Grade F)

ASTM A490 ASTM A194 or A563 ASTM F436 ASTM A36 In accordance with RGEE Specification CE-125 3.12.7 3.12.8

3. 13. 0

3. 13. 1 Welding Materials E70XX electrodes All replacement materials for the snubber hydrualic components shall be in accordance with RGGE Specification ME-245.

Mechanical Re uirements

.Re-installation of the snubber hydraulic tubing, reservoirs,

supports, and associated components shall be in accordance with RG&E Specification ME-245, the Design Criteria for EWR 2512, and design output documents.

3.13.1 3.14.0 3

~ 14. 1

3. 14. 2 No welding to the permanent steam generator supports is permitted.

Structural Re uirements The replacement steam generator support components shall be fabricated to the requirements of the AISC Manual of Steel Construction, 8th edition.

Consistent with UFSAR Section 3.9.3.2.2 and Reference 3.2.7, the replacement steam generator supports shall be evaluated in accordance with the ASME BGPV Code, 1974 Edition,Section III, Subsection NF and Appendix F.

3.15.0 3.15.1 3.16.0 3.16.1 H draulic Re uirements Not Applicable.

Chemist Re uirements Not Applicable.

Design Criteria DC-10034A EWR 10034 Page 54 Revision C

Date 09 26 94

t 1

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3. 17. 0 Electrical Re uirements

3. 17. 1

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3.18.0 3.18.1 Not Applicable.

La out and Arran ement Re uirements The size and shape of the new upper support (ring) shall match the existing upper support including the interface with the snubber and strut sets.

3.18.2 The design of the new support bracket assemblies shall be such that the interface with the steam generator support foot pads, the lower lateral supports, and the vertical support columns is identical.

3 '9.0 3.19.1 0 erational Re irements To preclude any possible adverse impacts on plant operation, the existing steam generator upper support (ring) and lower supports shall not be disconnected from the steam generator until the reactor has been defueled and the reactor cavity drained.

3. 19. 2 The new upper support (ring) and the new lower support components shall be installed and functional prior to refueling the reactor.

3. 19. 3

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Installation of the permanent steam generator supports shall be in accordance with Reference 3.2.7 and Drawing 33013-1979, and the Design Analysis for EWR 1483.

Instrumentation and Contxol Re irements Not Applicable.

Access and Administrative Control Re uirements 3.20.0 3.20.1 3.21.0 3.21.1 3.22.0 3.22.1 3.23.0 3.23.1 Not Applicable.

Redundanc Diversit and Se aration Re uirements Not Applicable'ailure Effects Re uirements new Design Criteria DC-10034A EWR 10034 Revision C

Page 55 Date 09 26 94 The new upper support (ring) and the new lower support components will be designed to satisfy the original design bases.

Replacement materials for the steam generator supports shall be compatible with the existing materials.

If a material substitution is

required, a reconciliation back to the original code or standard shall be performed to ensure the new materials meet or exceed the original material requirements.

The supports will not be installed until the reactor

I V

~4" P

rs

vessel has been defueled and the reactor cavity drained.

Consequently, no new failure effects need be considered.

3.24.0 3.24.1 Test Re uirements The new upper support bracket assemblies and pins shall be trial fit on the new steam generator support foot pads prior to installation in the steam generator compartment.

3.25.0 3.25.1 3.26.0 3.26.1 3.27.0 3.27.1 3.28.0 Accessibilit Maintenance Re air Re lacement Pum and Valve Test Pro ram and Znservice Ins ection Re uirements Not Applicable.

Personnel Re uirements Not Applicable.

Trans ortabilit Re uirements Not Applicable.

Fire Protection Re uirements 3.28.1

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The new steam generator support components are fabricated of steel.

No combustible materials are used.

Therefore, there will be no permanent increase in the probability or consequences of a fire and there will be no permanent increase in combustible loadings.

3.28.2 No existing fire barriers or fire barrier penetration seals will be affected and no new fire barriers will be installed.

3.28.3 The design and fabrication of the new steam generator support components shall use non-combustible and heat resistant materials whenever possible and to the extent practical.

3.28.4 The removal of the existing steam generator supports and the installation of the new steam generator support components shall not adversely affect the performance of any existing fire protection equipment.

3.28.5 3.28.6 Temporary fire protection provisions during the removal and installation of the steam generator supports shall be established.

An Appendix R Conformance Verification shall be performed to ensure that this modification does not Design Criteria DC-10034A EWR 10034 Page 56 Revision C

adversely impact existing Appendix R Complian methods.

3.29.0

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Handlin Re uirements 3.29.1 The existing upper support (ring) will remain attached to the steam generator.

Thus, the upper support will be removed from the containment when the steam generator is rigged through the containment opening.

3.29.2 Conversely, the new upper support (ring) will be installed on the steam generator outside the containment and rigged with the steam generator into place.

3.30.0 3.30.1 3.31.0 3.31.1 3.32.0 3.32.1 Public Safet Re uirements Not Applicable.

A licabilit Material requirements are identified in Section 3.12.0.

Personnel Safet Re irements Methods to remove and re-install the steam generator support components shall ensure that any occupational radiation exposures to personnel are As Low As Reasonably Achievable (ALARA).

3.32.2 The removal and re-installation of the permanent steam generator supports shall minimize the potential for personnel injury. All work shall be performed in accordance with OSHA requirements.

3.32.3 3.33.0 3.33.1 3.34.0 3.34.1 A Pre-ALARA Review Checklist will be completed prior to implementation of this work.

Environmental uglification Re uirements A 10CFR50.49 (EQ) Applicability checklist has been completed and is included as Attachment 1.

Uni ue Re uirements Installation of the permanent steam generator supports shall be in accordance with Reference 3.2.7, Drawing 33013-1979, and the Design Analysis for EWR 1483.

Design Criteria DC-10034A EWR 10034 Page 57 Revision C

Date 09 26 94

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SECTION 4.0 4.1.0 STEAM GENERATOR LOWER SUPPORT TEMPORARY RESTRAXNTS Summar Descri tion of the Desi n 4.1.1 Summary

4. 1'. 1. 1 The steam generator lower supports are pinned columns which will require stabilization before being disconnected from the steam generators.

4.1.1.2 During an. outage prior to the steam generator replacement (SGR)

outage, brackets will be installed (clamped) on the steam generator support columns and tube steel will be we$ ded to the existing primary manway access platforms.

Temporary stabilizing braces with integral turnbuckles will be loosely fit-up to the column brackets and tube steel.

The column brackets and tube steel will remain in place during plant operation; the stabilizing braces including turnbuckles will be removed prior to entry in a mode above Cold Shutdown.

4.1.1.3 During the SGR outage, the temporary stabilizing braces will be installed between the column brackets and tube steel.

The stabilizing braces may be loosely installed during defueling but the turnbuckles shall not be tightened until all fuel has been removed from the reactor and the reactor 'cavity drained.

The turnbuckles shall be tightened before the RCS piping is cut.

The turnbuckles shall be loosened and the stabilizing braces disconnected after 1/3 weld-out of the RCS piping is completed and before fuel reload.

The tube steel, column brackets, and stabilizing braces shall be removed prior to entry in a mode above Cold Shutdown.

4.1.1.4 This'section of the Design Criteria defines the requirements for the design and.analysis of the steam generator lower support temporary restraints.

4.1.1.5 Requirements for the design of temporary supports and/or restraints of the RCS piping are identified in Design Criteria Document DC-10034C.

4.1.2 Functions

4. 1.2. 1 The function of the steam generator lower supports is to provide deadweight support and seismic restraint (download and overturning restraint) of the steam generators during all modes of reactor operation.

The steam generator support columns are also designed to permit free thermal growth.

Design Criteria DC-10034A EWR 10034 Page 58 Revision C

Date 09 26 94

El t'4 h Wt r ~

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4. 1.2.2 The function of the lower support temporary restraints is to lock the lower steam generator support columns in place to facilitate installation of the new steam generators.

4.1.3 4.1.3.1 Performance Requirements The temporary lower support bracing must be adjustable and be installed and removed quickly during the SGR outage.

4.1.3.2 The temporary lower support restraints shall be capable of stabilizing the lower support columns when the RCS piping is cut from the steam generators.

4. 1.4

4. 1.4.1 Control Installation of the column brackets, tube steel, and stabilizing braces shall not impair the structural capacity of the containment vessel, the steam generator, the steam generator

supports, or the primary manway access platforms, nor shall it permanently affect the control of any existing systems, features or equipment.

4.1.5 Modes of Operation 4.1.5.1

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The column brackets will be installed in an outage prior to the SGR outage by clamping to the support columns.

'Tube steel will be welded. to the existing primary manway access platforms.

Hence, the column brackets and tube steel will be installed during plant operation.

4. 1.5.2 During the SGR outage, the stabilizing braces may be installed during defueling but the turnbuckles shall not be tightened until all fuel has been removed from the reactor and the reactor cavity drained.

The turnbuckles shall be tightened before the RCS piping is cut.

The turnbuckles shall be loosened after 1/3 weld-out of the RCS piping is complete.

The bracing and column brackets shall be removed prior to entry in a mode above Cold Shutdown.

4. 1.5.3 No work shall be performed on any system until it has been released for work by Operations.

4.2.0 4.2.1 Referenced Documents Bechtel Quality Assurance Program Plan for Rochester Gas s Electric Corporation, R.

E. Ginna Nuclear Power

Plant, Steam Generator Replacement

Project, Bechtel Job No.

22225.

Design Criteria DC-10034A EWR 10034 Page 59 Revision C

Date

'09 26 94

4.2.2 Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

4.2.3

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Special Processes

Manual, Ginna Nuclear Power Plant Steam Generator Replacement

Project, Bechtel Job No.

22225.

4.2.4 4.2.4.1 4.2.5 4.2.5.1 4.2.6 4.2.6.1 American Institute of Steel Construction (AISC)

Manual of Steel Construction, 9th edition American Welding Society (AWS)

D1.1-94, Structural Welding Code.

American Society for Testing and Materials (ASTM)

ASTM A36, Revision A-93, Standard Specification for Structural Steel.

4.2.6.2 ASTM A307, Revision A-93, Standard Specification for Carbon Steel Bolts and Studs 60,000 psi Tensile Strength.

4.2.6.3 ASTM A325-93, Standard Specification for Structural Bolts, Steel, Heat Treated 120/105 ksi Minimum Tensile Strength.

4.2.6.4

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ASTM A490-93, Standard Specification for Heat Treated Steel Structural Bolts, 150 ksi Tensile Strength.

4.2.6.5 ASTM A563-93, Standard Specification. for Carbon and Alloy Steel Nuts.

4.2.6.6 4.2.7 ASTM F436-93, Specification for Hardened Steel Washers.

RGsE Specification CE-125, Technical Requirements for Furnishing and Erecting Structural Steel, Revision 7.

4.3.0 4.3.1 Seismic Cate o

The steam generator lower supports are classified as safety-related, seismic Category I structures.

4.3.2 The steam generator primary manway access platforms.are classified as safety significant, seismic Category II/I structures.

4.3.3 The Design Criteria Revision C

The column brackets that will be installed during an outage prior to the SGR outage are classified as safety significant, seismic Category II/I and shall be analyzed to ensure they will not fail and adversely impact safety-related equipment during a seismic event.

column brackets will be clamped to the support DC-10034A EWR 10034 Page 60 Date 09 26 94

>>a 1~

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4.3.4 columns; no welding to the support columns shall be performed.

The tube steel that will be installed on the primary manway access platforms during an outage prior to the SGR outage is classified as safety significant, seismic Category II/I.

4.3.6 The remaining parts of the temporary lower support restraints will only be fully installed after all fuel has been removed from the reactor vessel.

Thus, the stabilizing braces including turnbuckles are classified as non-safety, non-seismic components.

4.3.5 Coatings for the column brackets and tube steel are classified as safety-related.

All other coatings are nonsafety-related.

4.4.0 4.4.1 4.5.0 4.5.1 4.6.0 4.6.1 4.7.0 ualit Grou Code Class Fluid S stems Onl The existing steam generator supports were designed and fabricated to the requirements of the AISC Manual of Steel Construction, 8th edition.

The supports were subsequently re-evaluated in accordance with the ASME BsPV Code, 1974 Edition,Section III, Subsection NF and Appendix F.

Electrical S stem Safet Classifications Not Applicable.

A Pro ram A licabilit The design and installation of the steam generator lower supports temp'orary restraints shall be performed in accordance with the Bechtel Quality Assurance Program Plan.

Codes Standards and Re lato Re uirements The following principal codes, standards, and regulatory requirements shall be used in the design of the temporary lower support restraints:

4.7.1 4.7.1.1 4.7.2 4.7.2.1 American Institute of Steel Construction (AISC)

Manual of Steel Construction, 9th edition American Welding Society (AWS)

D1.1-94, Structural Welding Code.

Design Criteria DC-10034A EWR 10034 Page 61 Revision C

4.8.0 4.8.1 Desi n Conditions The temporary lower support restraints shall be capable of stabilizing the lower support columns when the RCS piping is cut from the steam generators.

4.9.0 4.9.1 Load Conditions Load Combinations The column brackets, tube steel, and stabilizing braces shall be analyzed for the following load combination:

(1)

Dead load

+ live load Q

=

D

+ L In addition, the column brackets and tube steel shall be analyzed for the following load combination:

(2)

Dead load

+ live load

+ earthquake Q = D+ L+

E'HERE:

DE'ead load of the structure live load safe shutdown earthquake load load combination to be used with working stress design 4.9.2 Allowable Stresses (1)

Dead Load

+ live load Allowable stresses shall be as defined in the AISC Manual of Steel Construction, 9th edition.

(2)

Dead load

+ live load

+ earthquake Allowable stresses shall be as defined in the AISC Manual of Steel Construction, 9th edition, with an increase factor of 1.6 but shall not exceed 0.9 of the yield stress.

Design Criteria DC-10034A EWR 10034 Page 62 Revision C

Date 0

26 94

'I 4

J J

4.9.3 4.9.3.1 Loads on the Steam Generator Support Columns The steam generator lower support columns shall be evaluated for the additional weight of the brackets in accordance with UFSAR Section 3.9.3.2.2.

4.9.4 Loads on the Primary Manway Access Platforms 4.9.4.1 The steam generator primary manway platforms shall be evaluated for the addition of the temporary tube steel.

4.10.0 4.10.1 Environmental Conditions The temporary brackets, tube steel, and stabilizing braces shall be desigoed to withstand the worst.-case environmental conditions in the steam generator cubicles.

4.11.0 4.11.1 Interface Re uirements The temporary restraints of the lower supports will interface with the permanent lower supports and the primary manway access platform.

4.12.0 4.12.1 4.12.2 4.12.3 4.12.4 4.12.5 4.12.7 Material Re irements Structural steel Bolts Nuts Washers Coatings Welding Materials ASTM A36 ASTM A307, ASTM A325, or ASTM A490 ASTM A563 ASTM F436 In accordance with RGEE Specification CE-125 E70XX electrodes.

All welding performed in the field shall be in accordance with the Special Processes Manual.

4.13.0 4.13.1 Mechanical Re uirements No welding to the permanent steam generator supports is permitted.

4.14.0 4.14.1 Structural Re irements The column brackets and tube steel will be installed in an outage prior to the SGR outage and will remain in place until completion of the SGR.

The column brackets Design Criteria DC-10034A EWR 10034 Page 63 Revision C

Date 09 26 94

I

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and tube steel shall be designed to remain in place during a seismic event to ensure they do not fail and impact safety-related equipment.

4.14.2 The column brackets and tube steel shall not degrade the functioning or structural/seismic capacity of the structures to which they are attached.

4.14.3 The temporary stabilizing braces may be installed during the defueling phase of the work, but the turnbuckles shall not be tightened until after all fuel has been removed.

Similarly, the turnbuckles shall be loosened before refueling has begun.

4.15.0 4.15.1 4.16.0 4.16.1 4.17.0 4.17.1 4.18.0 4.18.1 H draulic Re uirements Not Applicable.

Chemistr Re irements Not Applicable.

Electrical Re uirements Not Applicable.

La out and Arran ement Re uirements The column brackets, tube steel, and stabilizing braces shall be located to best suit personnel and equipment access during the SGR outage while still providing adequate support of the lower support columns.

4.19.0 4.19.1 0 erational Re irements The column brackets and'tube steel will be installed during an outage prior to the SGR outage and are classified as safety significant, seismic Category II/I and shall be analyzed to ensure they will not fail and adversely impact safety-related equipment during a seismic event.

The column brackets and tube steel shall not degrade the functioning or structural/seismic capacity of the structures to which they are attached.

4. 19. 2 The remaining parts of the temporary lower support restraints will be installed and used during the SGR outage and will only be fully installed (tightened) after all fuel has been removed from the reactor vessel and the reactor cavity drained.

4.19.3 The column brackets, tube steel, and stabilizing braces shall be removed from the containment upon completion of the SGR outage.

Design Criteria DC-10034A EWR 10034 Page 64 Revision C

Date 0

26 94

4.20.0 Instrumentation and Control Re uirements 4.20.1

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4.21.0 4.21.1 4.22.0 4.22.1 4.23.0 4.23.1 Not Applicable.

Access and Administrative Control Re uirements Not Applicable.

Redundanc Diversit and Se aration Re uirements Not Applicable.

Failure Effects Re uirements The column brackets wi,ll be installed in an outage prior to the SGR outage and will remain in place until completion of the SGR.

The temporary brackets shall be designed to remain in place during a seismic event to ensure they do not fail and impact safety-related equipment.

4.23.2 The column brackets and tube steel shall be designed and installed so as not to degrade the structural/seismic capacity or the functioning of the structures to which they are attached.

4.23.3 The temporary stabilizing braces of the lower support columns will only be functional during the time that all fuel is removed from the reactor vessel and the reactor cavity drained.

Consequently, postulated failures of the temporary bracing during the SGR outage will have no adverse safety impacts.

"4.23.4 The column brackets, tube steel, and stabilizing braces shall be removed from the containment upon completion of the SGR outage.

4.24.0 4.24.1 4.25.0 Test Re irements Not Applicable.

Accessibilit Maintenance Re air Re lacement Pum and Valve Test Pro ram and Inservice Ins ection Re uirements 4.25.1 4.26.0 4.26.1 4.27.0 Not Applicable.

Personnel Re irements Not Applicable.

Trans ortabilit Re uirements Not Applicable.

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4.27.1 Design Criteria DC-10034A EWR 10034 Page 65 Revision C

Date 09 26 94

a

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4.28.0 4.28.1 Fire Protection Re uirements The lower support column restraints are temporary structures; therefore, there will be no permanent increase in the probability or consequences of a fire and there will be no permanent increase in combustible loadings.

4.28.2 No existing fire barriers or fire barrier penetration seals will be affected and no new fire barriers will be installed.

4.28.3 The design and fabrication of the lower support column temporary bracing shall use non-combustible and heat resistant materials whenever possible and to the extent practical.

4.28.4 This design and use of the lower support column temporary restraints shall not adversely affect the performance of any existing fire protection equipment.

4.28.5 Temporary fire protection provisions during the installation,

use, and removal of the temporary lower support restraints shall be established.

4.28.6 An Appendix R Conformance Verification shall be performed to ensure that this modification does not adversely impact existing Appendix R Compliance methods.

4.29.0 4.29.1 Handlin Re uirements No special handling requirements are needed for the brackets and braces except that they be handled in a safe manner according to the appropriate codes of standard practice.

4.29.2 The column brackets, tube steel, and stabilizing braces shall be pre-assembled to the extent practical to facilitate rapid installation in the steam generator cubicle.

4.30.0 4.30.1 4.31.0 4.31.1 Public Safet Re uirements Not Applicable.

A licabilit All weld end preparation,

welding, and NDE performed in the field shall be in accordance with the Special Processes Manual.

4. 31. 2 Material" requirements are identified in Section 4.12.0.

Design Criteria DC-10034A EWR 10034 Page 66 Revision C

Date 09 26 94

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4.31.3 The temporary support column restraints shall remain functional without damage to surrounding equipment until removal after SGR completion.

4.32.0 4.32.1 Personnel Safet Re uirements The steam generator lower support column temporary restraints shall be designed to ensure that any occupational radiation exposures to personnel are As Low As Reasonably Achievable (ALARA).

4.32.2 The design, installation,

use, and removal of the temporary brackets and bracing shall'inimize the potential for personnel injury. All work shall be performed in accordance with OSHA requirements.

4.32.3 A Pre-ALARA Review Checklist will be completed prior to implementation of this work.

4.33.0 4.33.1 Environmental uglification Re uirements A 10CFR50.49 (EQ) Applicability checklist has been completed and is included as Attachment 1.

4.34.0 Uni ue Re uirements Not Applicable.

Design Criteria DC-10034A EWR 10034 Page 67 Revision C

Date 0

26 94

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SECTION 5.0 REACTOR CAVITY DECKING 5.1.0

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Summar Descri tion of the Desi n 5.1.1 Summary 5.1.1.1 5.1. 1.2 To provide a general work and construction laydown area inside the containment, a section of the area over the reactor cavity will be temporarily covered with steel decking.

This decking will be installed in pre-fabricated panels after all fuel is removed from the containment, the reactor cavity drained, and necessary isolation of the spent fuel pool and supporting containment systems iS achieved.

The reactor cavity decking shall be removed prior to fuel load.

The purpose of this section of the Design Criteria is to define the structural and analysis requirements for the reactor cavity decking.

5.1.2 5.1.2.1 Functions The function of the reactor cavity decking is to support personnel, equipment, and material laydown during the steam generator replacement (SGR) outage.

5.1.2.2

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An additional function of the reactor cavity decking is to protect the reactor cavity from dropped objects.

5.1.3 Performance Requirements 5.1.3.1 The performance requirements of the reactor cavity decking are to provide supplemental floor area for equipment and personnel during the SGR outage.

5. 1.3.2 The reactor cavity decking design live load shall be 200 psf.

5.1.4 5.1.4.1 Control The installation, use,.and removal of. the reactor cavity decking shall not impair the structural capacity of the containment, nor shall it adversely affect the control of any existing systems, features or equipment.

Modes of Operation

5. 1.5 5.1.5.1 dec Design Criteria Revision C

The reactor cavity decking shall be installed,

used, and removed with the reactor shut down, all fuel removed from the containment, the reactor cavity

drained, and all necessary isolation of the spent fuel pool and supporting containment systems achieved.

The king shall be capable of performing its function DC-10034A EWR 10034 Page 68

5.2.0

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l throughout the SGR outage.

The reactor cavity decking shall be removed prior to fuel load.

Referenced Documents 5.2. 1 5.2.2 5.2.3 5.2.3.1 5.2.4 5.2.4.1 5.2.5 5.2.5.1 5.2.6 5.2.6.1

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5.2.6.2 5.2.6.3 5.2.6.4 5.2.6.5 5.2.6.6 5.2.6.7 Bechtel Quality Assurance Program Plan for Rochester Gas

& Electric Corporation, R.

E. Ginna Nuclear Power

Plant, Steam Generator Replacement

Project, Bechtel Job No.

22225.

Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

American Concrete Institute (ACI) 318-63, Building Code Requirements for Reinforced Concrete and Commentary.

American Institute of Steel Construction (AISC)

Manual of Steel Construction, 9th edition American Welding Society (AWS)

D1.1-94, Structural Welding Code.

American Society for Testing and Materials (ASTM)

ASTM A36, Revision A-93, Standard Specification for Structural Steel.

ASTM A307, Revision A-93, Standard Specification for Carbon Steel Bolts and Studs 60,000 psi Tensile Strength.

ASTM A325-93, Standard Specification for Structural Bolts, Steel, Heat Treated 120/105 ksi Minimum Tensile Strength.

ASTM A446-93, Standard Specification for Sheet Steel, Zinc-Coated (Galv)

By the Hot Dip Process, Structural (Physical Quality).

ASTM A490-93, Standard Specification for Heat Treated Steel Structural Bolts, 150 ksi Tensile Strength.

ASTM A563-93, Standard Specification for Carbon and Alloy Steel Nuts.

ASTM A569, Revision A-91, Specification for Steel, Carbon

(.15 Max Percent)

Hot-Rolled Sheet and Strip, Commercial Quality.

Design Criteria DC-10034A EWR 10034 Page 69 Revision C

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5.2.6.8 ASTM A659-92, Specification for Steel, Carbon

(.16 Min to.25 Max Percent)

Hot-Rolled Sheet and Strip, Commercial Quality.

5. 2.7 RGSE Specification CE-125, Technical Requirements for Furnishing and Erecting Structural Steel, Revision 7.

5.3.0 5.3.1 Seismic Cate or The reactor cavity decking is classified as a non-

safety, non-seismic structure.

5.3.2 5.3.3 The containment supporting structures are classified as safety-related, seismic Category I structures.

Because the reactor cavity decking will not be in the containment during plant operation, coatings are classified as nonsafety-related.

5.4.0 5.4.1 5.5.0 5.5.1 5.6.0 5.6.1 ualit Grou Code Class Fluid S stems Onl Not Applicable.

Electrical S stem Safet Classifications Not Applicable.

A Pro ram A licabilit The design, installation,

use, and removal of the temporary reactor cavity decking shall be performed in accordance with the Bechtel Quality Assurance Program Plan.

5.7.0 Codes Standards and Re lator Re uirements The following principal codes, standards, and regulatory requirements shall be used in the design of the reactor cavity decking or in the evaluation of the supporting structure:

5.7.1 5.7. 1. 1 5.7.2 5.7.2.1 American Concrete Institute (ACI) 318-63, Building Code Requirements for Reinforced Concrete and Commentary.

American Institute of Steel Construction (AlSC)

Manual of Steel Construction, 9th edition 5.7.3 5.7.3.1 American Welding Society (AWS)

I D1.1-94, Structural Welding Code.

Design Criteria DC-10034A EWR 10034 Page 70 Revision C

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5.8.0t 5.8.1 Desi n Conditions The reactor cavity decking design live load shall be 200 psf.

5.8.2 The reactor cavity decking shall be considered as live load on the supporting structure.

5.9.0 5.9.1 5.9.1.1 Load Conditions Load Combinations The reactor cavity decking shall be analyzed for the following load combination:

Q

=

D + L WHERE:

D =

dead load of the structure and equipment L =

live load of 200 psf Q

=

load combination to be used with working stress design The containment supporting structures shall be evaluated for the following load combination:

U =

1

~ 5D + 1.8L WHERE:

D =

Dead load of structure L =

Load imposed on supporting structure from reactor cavity decking U =

Load combination to be used with ultimate strength design 5.9.2 5.9.2.1 Allowable Stresses Allowable stresses for the design of the reactor cavity decking shall be as defined in the AISC Manual of Steel Construction, 9th edition.

5.9.2.2 Allowable stresses for the evaluation of the containment supporting structures shall be as defined in ACI 318-63.

5.10.0 Environmental Conditions Not Applicable.

t

5. 10. 1 Design Criteria DC-10034A EWR 10034 Page 71 Revision C

Date 09 26 94

5.11.0

5. 11. 1 Interface Re uirements The reactor cavity decking will be installed over sections of the reactor cavity after the plant has been shut down, all fuel removed from the containment, the reactor cavity drained, and all necessary isolation of the spent fuel pool and supporting containment systems is achieved.

The reactor cavity decking shall be removed prior to fuel load.

5. 11. 2 The supporting containment structure shall be evaluated for the addition of the cavity decking as a live load.

5.12.0 5.12.1 5.12.2

5. 12.3 5.12.4 Material Re uirements Structural steel Bolts Nuts Sheeting/decking ASTM A36 ASTM A307, ASTM A325, or ASTM A490 ASTM A563 ASTM A446, A569 or ASTM A659 5.12.5 5.12.6

~

~

5.13.0

5. 13. 1 5.14.0 Coatings

'Weld'ing Materials Mechanical Re uirements Not Applicable.

Structural Re irements In accordance with RG6E Specification CE-125 E70XX electrodes 5.14.1 The reactor cavity decking shall be capable of supporting the required loads without jeopardizing the structural integrity of any component or structure inside the containment.

5. 15. 0 5.15.1 5.16.0 5.16.1

5. 17. 0 5.17.1 H draulic Re uirements Not Applicable.

Chemist Re uirements Not Applicable.

Electrical Re irements Not Applicable.

Design Criteria DC-10034A EWR 10034 Page 72 Revision C

Date 09 26 94

5.18.0 5.18.1 La out and Arran ement Re uirements The layout and arrangement of the reactor decking shall best suit access and laydown requirements during the SGR outage.

5.19.0 5.19.1 5.20.0 5.20.1 5..21. 0 5.21.1 5.22.0 5.22.1 5.23.0 5.23.1 0 erational Re uirements Not Applicable.

Instrumentation and Control Re uirements Not Applicable.

Access and Administra ive Control Re uirements Not Applicable.

Redundanc Diversit and Se aration Re uirements Not Applicable.

Failure Effects Re uirements The reactor cavity decking will be installed in pre-fabricated panels after all fuel is removed from the containment, the reactor cavity drained, and all necessary isolation of the spent fuel pool and supporting containment systems is achieved.

The decking will be removed prior to fuel load.

Consequently, postulated failures of the cavity decking will have no adverse impacts on plant safety.

5.24.0 5.24.1 5.25.0 5.25.1 5.26.0 5.26.1 5.27.0 5.27.1 Test Re uirements Not Applicable.

Accessibilit Maintenance Re air Re lacement Pum and Valve Test Pro ram and Inservice Ins 'ection Re uirements Not Applicable.

Personnel Re uirements Not Applicable.

Trans ortabilit Re uirements Not Applicable.

Design Criteria DC-10034A EWR 10034 Page 73 Revision C

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5.28.0 5.28.1 Fire Protection Re irements The reactor cavity decking is a temporary structure; therefore, there will be no permanent increase in the probability or consequences of a fire and there will be no permanent increase in combustible loadings.

5.28.2 No existing fire barriers or fire barrier penetration seals will be affected and no new fire barriers will be installed.

5.28.3 The design and fabrication of the reactor cavity decking shall use non-combustible and heat resistant materials whenever possible and to the extent practical.

5.28.4 This design and use of the reactor cavity decking shall not adversely affect the performance of any existing fire protection equipment.

5. 28. 5-Temporary fire protection provisions during the installation,

use, and removal of the reactor cavity decking shall be established.

5.28.6 An Appendix R Conformance Verification shall be performed to ensure that this modification does not adversely impact existing Appendix R Compliance methods.

5.29.0 Handlin Re uirements

5. 29. 1 No special handling requirements are needed for the reactor cavity decking except that it be handled in a safe manner according to the appropriate codes of standard practice and the safe load, paths defined in RGsE Administrative Procedures A-1305 and A-1305.1.

5.29.2 The materials required for the reactor cavity decking shall be pre-assembled in pieces that will fit through the equipment hatch to facilitate rapid installation and removal.

5.30.0 5.30.1 5.31.0 5.31.1 Public Safet Re uirements Not Applicable.

A licabilit Naterial requirements are identified in Section 5.12.0.

Design Criteria DC-10034A EWR 10034 Page 74 Revision C

5.32.0 5.32.1 Personnel Safet Re uirements The reactor cavity decking shall be designed to ensure that any occupational radiation exposures to personnel are As Low As Reasonably Achievable (ALARA).

5.32.2 The reactor cavity decking shall be fabricated and coated to facilitate decontamination.

5.32.3 The design, installation,

use, and removal of the temporary structures shall minimize the potential for personnel injury. All work shall be performed in accordance with OSHA reguirements.

5.32.4 5.33.0 5.33.1 A Pre-ALARA Review Checklist will be completed prior to the implementation of this work.

Environmental uglification Re uirements A 10CFR50.49 (EQ) Applicability checklist has been completed and is included as Attachment 1.

5.34.0 5.34.1 Uni ue Re irements Not Applicable.

Design Criteria DC-10034A EWR 10034 Page

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SECTION 6.0 6.1.0 6.1.1 INTERFERING COMMODITIES Summar Descri tion of the Desi n Summary 6.1.1.1 To facilitate the removal and replacement of the steam generators, two openings will be made in the containment vessel dome (see Section 1.0 of this document).

For these openings to be made, sections of the containment spray piping, containment HVAC duct, and painter's trolley must be temporarily removed from the inside surface of the containment dome.

6.1.1.2 Containment Spray Piping

6. 1. 1. 2. 1 The containment spray piping ring at Elevation 372'-8" is an interference to the containment roof openings over both steam generators (SGs)

A and B.

6. 1. 1. 2. 2 6.1.1.2.3 During the SGR outage, two sections of the containment spray ring, approximately 21-feet-long over SG A and 46-feet-long over SG B, must be temporarily removed from the inside of the containment dome.

It will also be necessary to remove containment spray piping supports over SG A and SG B.

The spray piping will be mechanically cut and re-connected using butt welds.

As an option to cutting and performing butt, welded connections to the containment spray piping during the SGR outage, flanged connections may be installed.

In an outage prior to the SGR outage, flanges may be installed on the spray ring piping.

By installing these flanges in a prior outage, the construction effort to remove and re-install the spray piping during the SGR outage will be simplified and minimized.

6. 1.1.3 6.1.1.3.1 Containment Recirculation and Cooling System Duct I

The containment recirculation and cooling system duct running vertically up the inner containment wall to the top of the dome is an interference to the containment roof opening over steam generator (SG)

B.

One section of circular ducting approximately 20-feet-long will be disassembled and temporarily removed from the inside of the containment dome.

6.1.1.3.2 The section of ducting to be removed is 28" in diameter at the lowest elevation being removed and reduces down to 24" for the majority of the span being removed (see Drawing 33013-1863).

It will also be necessary to remove the ventilation ducting supports located in the vicinity of the containment roof opening over SG B.

Design Criteria DC-10034A EWR 10034 Page 76 Revision C

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6.11,3.3 Two flexible duct connections containing asbe aterial will be also be removed and discarded as art m

p of this modification.

Two new flexible duct connections will be installed.

Removal and replacement of the existing flexible connections may be performed in an outage prior to the SGR or during the SGR outage.

6.1.1.4 6.1 '.4.1 Painter's Trolley Rails The four painter's trolley rails attached to the inside of the containment dome are located at El. 345'-2", El.

361'-11", El. 377'-10",

and El. 382'-7".

Only the rail at El. 377'-10" interferes with the dome openings.

A section of the rail at El. 377'-10" must be temporarily removed at the opening over SG A and the opening over SG B.

6. 1.2

6. 1.2. 1

6. 1.2. 1. 1 Functions Containment Spray System The containment spray system, in conjunction with the containment cooling system and the emergency core cooling system, is designed to.remove sufficient heat from the containment atmosphere following an accident condition to maintain the containment pressure below structural limits.

The containment spray system is also capable of reducing the iodine and particulate fission product inventories in the containment atmosphere such that the offsite radiation exposure resulting from a loss-of-coolant accident is within the guidelines established by 10 CFR 100.,

6.1.2.2 6.1.2.2.1 Containment Recirculation and Cooling System The containment recirculation and cooling system provides ventilation, filtration and cooling of the containment structure during all modes of reactor operation.

6.1.2.3 6.1.2.3.1 Painter's Trolley Rail The painter's trolley rail provides a means of access to the containment dome for maintenance activities.

6.1.3 6.1.3.1 6.1.3.1.1 Performance Requirements Containment Spray System The containment spray system is designed to operate over an extended time period following a primary coolant system failure as required to restore and maintain containment conditions at near atmospheric ssure.

The spray system delivers at least 2400 gpm Design Criteria DC-10034A EWR 10034 Page 77 Revision C

Date 09 26 94

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f borated water into the containment building.

Either 0of two subsystems containing a pump and associated valving and spray headers is independently capable of delivering one half of this flow, or 1200 gpm.

6.1.3.2 Containment Recirculation and Cooling System 6.1.3.2.1 The portion of the ventilation ducting being removed is designed to remove the normal heat loss from the equipment and piping in the reactor containment during plant operation and maintain a normal ambient temperature of about 120'F, 50% relative humidity.

During periods of reactor shutdown, this system operates to provide a minimum containment ambient temperature of 50 F.

.This system also provides sufficient air circulation and filtering throughout all containment areas to permit safe and continuous access to the reactor containment within two (2) hours after reactor shutdown.

6.1.3.3 6.1.3.3.1 Painter's Trolley Rail The painter's trolley rail is des'igned to facilitate personnel access to the containment dome and support equipment during unit shutdown.

6.,1. 4

6. 1.4. 1 Control Modifications to the control functions associated with the containment spray system or the containment recirculation and cooling system will not be performed.

No'ontrol functions are affected by the planned modifications to the painter's trolley rails.

6.1.5 6.1.5.1 Modes of Operation The work performed to this design criteria shall not affect the modes of operation of any system or component.

6.1.5.2 Removal of the containment spray piping and supports as well as the ductwork and supports associated with the containment recirculation and cooling system shall only be performed once the reactor has entered a Cold Shutdown condition.

Reinstallation of the piping/ductwork and supports, including applicable system testing, shall be performed prior to entering a

mode above Cold Shutdown.

6.1.5.3 The painter's trolley rail may be removed once the reactor has entered the Cold Shutdown condition.

6.1.5.4 No work shall be performed on any system until it has been released for work by Operations.

I Design Criteria DC-10034A EWR 10034 Page 78 Revision C

Date 09 26 94

6.2.0 6.2.1 6.2.2 6.2.3 Referenced Documents Bechtel Quality Assurance Program Plan for Rochester Gas

& Electric Corporation, R.

E. Ginna Nuclear Power

Plant, Steam Generator Replacement

Project, Bechtel Job No.

22225.

Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

Special Processes

Manual, Ginna Nuclear Power Plant Steam Generator Replacement

Project, Bechtel Job No.

22225.

6.2.3 6.2.3.1 6.2.4 6.2.4.1 6.2.5 6.2.6 6.2.7 6.2.8 6.2.9 American Institute of.Steel Construction (AISC)

Manual of Steel Construction, 9th edition.

American Welding Society (AWS)

D1.1-94, Structural Welding Code.

ASME Section III, 1986 Edition, Boiler & Pressure Vessel Code, Division 1.

ASA B31.1-1955, Code for Pressure Piping ASA B16.5-1961.

EWR 2512, Seismic Upgrade Program.

Design Criteria, Seismic Upgrade

Program, EWR 2512, Reference 5.

6.2.10 6.2.11 6.2.12

6. 2. 13 6.2. 14 6.2.15 Westinghouse E-Spec.

G-569866, "Line Specifications (Nuclear Plant)," Revision 2, dated 12/23/66.

Gilbert Associates, Inc., Reactor Building Seismic Analysis, Floor Response

Spectra, Ginna Station, Seismic Upgrading Program, EWR 2512, October 12, 1979.

Design Criteria, Replacement Containment Recirculating Fan Cooler Coil Installation, EWR 5275, Revision 0.

RG&E Specification ME-121, Technical Specification for the Fabrication and Installation of Seismic Category I Pipe Supports.

RG&E Specification CE-125, Technical Requirements for Furnishing and Erecting Structural Steel, Revision 7.

RG&E Specification SP-5342, Heating, Ventilation and Cooling Systems, through Addendum B.

Design Criteria DC-10034A EWR 10034 Page 79 Revision C

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I I

6.2.16 Drawing 9-5159, Sheet 47, Location of Painter'rolley Rails 6.2.17 Drawing 9-5159, Sheet 48, Painter's Trolley Rail Details 6.2.18 6.2.19 6.2.20 6.2.21 6.2.22 6.2.23 6.2.24 6.2.26 6.3.0 6.3.1

,6. 3. 2 6.3.3

6. 3.4 6.3.5 Drawing 33013-1261, Containment Spray Piping (SI)

Drawing C-381-359, Sheets 1 through 5, Containment Spray Piping Drawing D-304-641, Containment

Spray, Plans and Sections Drawing D-521-061, Reactor Vessel Containment, Misc.

Supports for Vent Ducts and Spray Piping Drawing 33013-1863, Containment HVAC System Drawing D-118-025, Reactor Containment Vessel Heating, Ventilation and Filtering Part 56 of Title 12 of the Official,Compilation of

Codes, Rules and Regulations of the State of New York, 12 NYCRR Part 56, Asbestos.

Title 29 of the Code of Federal Regulations, Parts 1926 and 1910, Occupational Safety a Health Administration (OSHA) Regulations (29 CFR 1926/1910).

Title 40 of the Code of Federal Regulations, Part 61, Environmental Protection Agency (EPA) Regulations (40 CFR 61).

Seismic Cate o

The containment dome and liner plate are classified as safety-related, seismic Category I.

The containment spray piping and supports are classified as safety-related, seismic Category I.

The containment recirculation and cooling system is a safety-related, seismic Category I system.

The painter's trolley rails are classified as safety significant, seismic Category II/I. All work associated with the removal and re-installation of the painter's trolley rails is safety significant with the exception of the weld to the containment dome liner plate which is safety-related.

Coatings are classified as safety-related.

Design Criteria DC"10034A EWR 10034 Page 80 Revision C

Date 09 26 94

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6.4.0 6.4.1 6.5.0 6.5.1 6.6.0 6.6.1 6.7.0 ualit Grou Code Class Fluid S stems Onl The containment spray system is Quality Group B, as defined in Regulatory Guide 1.26 and ASME Section II1, Class 2, piping.

Electrical S stem Safet Classifications Not Applicable.

A Pro ram A licabilit All modifications to the containment spray piping, HVAC duct, painter's trolley and associated supports shall be performed in accordance with the Bechtel Quality Assurance Program Plan.

Codes Standards and Re lato Re uirements 6.7. 1

6. 7. 2.

6.7.2.1 6'.3 6.7.3.1 6.7.4 6.7.4.1 6.7.5 6.7.6 6.7.7 6.7. 8 The following principal codes, standards, and regulatory reguirements shall be used in the modification of the systems addressed in Section 6.0 of this design criteria:

Ginna Station Updated Final Safety Analysis Report (UFSAR), Revision 10, 12/93.

American Institute of Steel Construction,(AISC)

Manual of Steel Construction, 9th edition American Welding Society (AWS)

D1.1-94, Structural Welding Code.

American Society of Mechanical'Engineers ASME Section III, 1986 Edition, Boiler s Pressure Vessel Code, Division 1.

ASA B31.1-1955, Code for Pressure Piping Design Criteria, Seismic Upgrade

Program, EWR 2512, Reference 5, Dated 4/11/89.

Westinghouse E-Spec.

G-569866, "Line Specifications (Nuclear Plant)," Revision 2, dated 12/23/66.

Gilbert Associates, Inc., Reactor Building Seismic Analysis, Floor Response

Spectra, Ginna Station, Seismic Upgrading Program, EWR 2512, October 12, 1979.

Design Criteria DC-10034A EWR 10034 Page 81 Revision C

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6.8.0 6.8.1 Desi n Conditions Containment Spray Piping 6.8.1.1 The containment spray piping that must be removed is ASME Section III, Class 2 and was designed in accordance with ASA B31.1-1955 and ASA B16.5-1961.

The spray piping line specification is 301R.

6.8. 1.2 The standard design conditions for the containment spray system are as follows:

6.8.2 6.8.2.1 Design Temperature:

300'psig Design Temperature:

300 'F Containment Recirculation and Cooling System The containment recirculation and cooling system ductwork shall be designed to withstand a maximum of 30" wg internal positive pressure with 0 psig external to the duct and maximum 2 psig externally -applied pressure with 0 psig internal duct pressure per SP-5342.

6.8.2.2 The flexible duct connections shall be fiberglass cloth with flame resistant, silicone impregnated cloth.

Mounting flanges in flexible connections shall be, suitable for welding.

Flexible duct connections shall be capable of handling 2." axial movement; 1" lateral movement, differential pressure of 1.5 psig positive internal pressure or 2 psig positive external pressure per SP-5342.

6.8.3 6.8.3.1 Painter's Trolley Monorail Design conditions for the painter's trolley monorail are defined in the Design Criteria for EWR 2512, Seismic Upgrade Program.

6.9.0 6.9.1 Load Conditions Design-of piping, ductwork, supports and structures shall consider the applicable

seismic, deadweight,

pressure, and thermal loads specified in Section 6.8.

Seismic loading conditions shall be as specified in the Design Criteria for EWR 2512, Seismic Upgrade Program.

6.9.2 If flanges are added to the containment spray piping, the piping, supports, and liner plate shall be evaluated for the increased loading and determined to be acceptable.

Design Criteria DC-10034A EWR 10034 Page 82 Revision C

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6.10.0 6.10.1 Environmental Conditions As specified in the Design Criteria for EWR 5275, for installation in containment, all equipment shall be designed to withstand the following environmental conditions without impairing operability:

Normal Accident Ambient Temperature

('F)40-120 Ambient Pressure (psig) 0-3 Relative Humidity

(%)10-100 Radiation Dose Rate/Dose

<1 rad/hr 286 60 100 1.43E7 rad (gamma) 2.07E8 rad (beta) 6.11.0

6. 11. 1 6.12.0 6.12.1 6.12.2 Under post-accident conditions, the exterior of the piping, ductwork and rails shall be subjected to a chemical spray containing 3,000 ppm boric acid and sodium hydroxide to create a solution pH of 8-10.

Interface Re uirements The modifications associated with the interfering commodities shall maintain the structural and functional integrity of the affected components and shall not degrade the ability of the affected components to perform according to their design basis.

Material Re uirements Replacement materials for the interfering commodities shall be compatible with the existing materials.

Xf a material substitution is required, a

reconciliation back to the original code or standard shall be performed to ensure the new materials. meet or exceed the original material requirements.

6.12.3 6.12.3.1 6.12.3.2 6.12.3.3 Containment Spray Piping Piping Flanges Flange Bolting ASTM/ASME A/SA 312 TP304 ASTM/ASME A/SA 182 F304, Class 300 Raised face, slip-on ASTM/ASME A/SA 193, Grade B7 Bolts ASTM/ASME A/SA 194, Grade 2H Nuts Design Criteria DC-10034A EWR 10034 Page 83 Revision C

Date 0

26 94

<<W.

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6. 12. 3. 4 Flange Gasket Spiral Wound Stainless Steel With Graphite Filler 6 '2.3.5 Supports

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6.12.3.6 Support Bolting 6.12.3.7 Support Coating 6.12.3.8 Welding Materials ASTM A515 Grade 60 ASTM A193, Grade B7 Bolts ASTM A194, Grade 2H Nuts ASTM F436 Washers In accordance with RG6E Specification CE-125.

All welding performed in the field shall be in accordance with the Special Processes Manual.

6.12.4 Containment HVAC Duct 6.12.4.1 Flexi'ble Connections 6.12.4.2 Supports 6.12.4.3 Bolts, Washers, Nuts Johns-Manville F-350 or equivalent ASTM A36 ASTM A325 Bolts ASTM A563 Nuts ASTM F436 Washers 6.12.4.4 Support Coatings

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6.12.4.5 Welding Materials In accordance with RGSE Specification CE-125.

All welding performed in the field shall be in accordance with the Special Processes Manual.

6.12.5 Painter's Trolley 6.12.5.1 Monorail 6.12.5.2 Supports 6.12.5.2 Coatings 6.12:5.3 Welding Materials ASTM A36 ASTM A516 Grade 60 In accordance with RG&E Specification CE-125.

All welding performed in the field shall be in accordance with the Special Processes Manual.

Design Criteria DC-10034A EWR 10034 Page 84 Revision C

Date 09 26 94

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6. 13.0 6.13.1 Mechanical Re uirements All piping and support modifications shall be consistent with the Design Criteria for EWR 2512.

6.13.2 The mechanical requirements for the containment spray system and containment HVAC system are not affected by the work performed under this design criteria.

6.13.3 Re-installation of the containment spray piping and supports shall satisfy the requirements of RGSE Specification ME-121.

The reinstalled configuration shall be identical to the existing configuration as shown on Drawings 33013-1261, C-381-359, D-304-641, and D-521-061.

Weld end preparation,

welding, and NDE shall be in accordance with the Special Processes Manual.

6. 13.4 Re-installation of the containment HVAC ductwork and associated supports shall satisfy the requirements of RGEE Specification SP-5342.

The reinstalled configuration shall be identical to the existing configuration as shown on Drawings D-521-061, 33013-

1863, and D-118-025.

Weld end preparation,

welding, and NDE shall be in accordance with the Special Processes Manual.

6. 14. 0 6.14.1 Structural Re uirements The removed sections of painter's trolley rails at El.

377'-10" shall.be reinstalled in accordance with the original design as depicted on CBSI Drawing 9-5159, Sheets 47 E 48.

No changes to the design or design bases will be made.

Therefore, no reanalysis is required.

6. 14.2 Any changes in supports or attachments to the liner plate shall be evaluated and determined to be acceptable.

6. 15.0

6. 15. 1 6.16.0

6. 16. 1 6.16.0 6.16.1 H draulic Re uirements Not Applicable.

Chemistr Re irements Not Applicable.

Electrical Re uirements Not Applicable.

Design Criteria DC-10034A EWR 10034 Page 85 Revision C

Date 0

26 94

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6.18.0 6.18.1 La out and Arran ement Re uirements Upon completion of the modifications associated with the interfering commodities, the layout of the affected components will be identical to the pre-modification configuration.

6.19.0 6.19.1 0 erational Re uirements The modifications to the interfering commodities shall not unacceptably change the functional capabilities of the containment spray system, the containment recirculation and cooling system or the painter' rails.

6.20.0 6.20.1 6.21.0 6.21.1 6.22.0 6.22.1 6.23.0 6.23 '

Instrumentation and Control Re uirements Not Applicable.

Access and Administrative Control Re uirements Not Applicable.

Redundanc Diversit and Se aration Re irements Not Applicable.

Failure Effects Re uirements The modifications to the containment

spray, containment recirculation and cooling system and the painter' trolley. rail shall be analyzed to demonstrate that the affected components will remain functional under all applicable design basis events.

6.24.0 6.24.1 6.24.1.1 Test Re irements Containment Spray System Test results indicating complete nozzle availability shall be observed prior to entering a mode above Cold Shutdown.

Prior to the test, spray nozzle orientation shall be verified to be in accordance with Drawing D-304-641.

Testing of containment spray shall satisfy the requirements specified in UFSAR Section 6.2.2.2.5 and shall consist of either injection of smoke or low

pressure, heated air (monitored by thermographs) through the existing test connection.

Containment Recirculation and Cooling System 6.24.2 6.24.2.1 sha Design Criteria Revision C

When the HVAC system has been fully restored, the containment recirculation cooling and filtration system ll be tested to verify system flow.

Test results DC-10034A EWR 10034 Page 86 Date 09 26 94

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labilit shall be observed indicating system avai Y

prior to entering a mode above Cold Shutdown.

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6.24.3 Painter's Trolley Rail 6.24.3.1 There are no specific testing requirements associated with this modification.

6.25.0 6.25.1 6.26.0 6.26.1 Accessibilit Maintenance Re air Re lacement Pum and Valve Test Pro ram and Inservice Ins ection Re uirements Not Applicable.

Personnel Re uirements The flexible duct connections to be removed contain asbestos material.

Appropriate precautions shall be taken during the handling of these flexible duct connections to preclude asbestos contamination.

6.26.0 6.26.1 6.28.0 Trans ortabilit Re uirements

'Not Applicable.

Fire Protection Re uirements 6.28.1

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The systems affected by removal/restoration of the interfering commodities have no fire protection function.

In addition, the activities associated with the modifications will be performed in accordance with existing fire protection requirements; therefore, there will be no permanent increase in the probability or consequences of a fire and there will be no permanent increase in combustible loadings.

6. 28. 2 No existing fire barriers or fire barrier penetration seals will be affected and no new fire barriers will be

'nstalled.

6.28.3 The modification shall use non-combustible and heat resistant materials whenever possible and to the extent practical.

6 '8.4 The modifications associated with removal/reinstallation of the interfering commodities shall not adversely affect the performance of any existing fire protection equipment.

6.28.5 Temporary fire protection provisions shall be established, as required, during the removal/restoration of interfering commodities.

Design Criteria DC-10034A EWR 10034 Page 87 Revision C

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An Appendix R Conformance Verification shall be performed to ensure that this modification does not adversely impact existing Appendix R Compliance methods.

6.29.0 6.29.1 6.29.1.1 6.29.2 I

6.29.2.1 Handlin Re uirements Containment Spray System The spray piping will be mechanically cut and the open pipe ends will be prepared for reinstallation and temporarily sealed to prevent the entry of debris into the system.

During the cutting process, appropriate measures shall be taken to preclude the introduction of debris into the piping.

The removed sections of containment spray piping will be prepared for reinstallation and shall be properly stored during the SGR outage to ensure the condition of the piping and/or spray nozzles is not degraded.

Containment Recirculation and Cooling System Considering the ventilation system's, filtering function, special precautions shall be taken to reduce the potential for radioactive contamination during disassembly of the ventilation ducting.

The open ends of ventilation ducting will be temporarily sealed to prevent the entry of debris into the system and allow the continued operation of the remainder of the containment recirculation and cooling system during the SGR outage.

6.29.2.2 The removed sections of ventilation ducting shall be properly stored during the SGR outage'to ensure the condition of the duct work is not degraded.

6. 29. 2. 3 The flexible duct connections to be removed contain asbestos material.

Appropriate precautions shall be taken during the handling of these flexible duct, connections to preclude asbestos contamination.

6.29.3 6.29.3.1 6.30.0 6.30.1 6.31.0 6.31.1 Painter's Trolley Rail No specific handling requirements are applicable.

Public Safet Re uirements Not Applicable.

A licabilit Pro Design Criteria Revision C

All weld end preparation,

welding, and NDE performed in the field shall be in accordance with the Special cesses Manual.

DC-10034A EWR 10034 Page 88 Date 09 26 94

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6. 31.2 6.32.0 Material requirements are identified in Section 6.12.0.

Personnel Safet Re irements 6.32.1 Modifications to the containment spray piping, HVAC

duct, and painter's trolley shall be designed to ensure that any occupational radiation exposures to personnel are As Low As Reasonably Achievable (ALARA).

6. 32.2 The flexible duct connections to be removed contain asbestos material.

All work associated with removal of asbestos-bearing materials shall be performed by a certified asbestos removal contractor and certified personnel in accordance with all federal and state regulations governing. asbestos control and removal.

These regulations include 12 NYCRR Part 56, 29 CFR 1926/1910, and 40 CFR 61.

6.32.3 All removed asbestos-containing materials shall be properly bagged,

labeled, and sealed into approved containers.

6.32.4 All controls necessary to prevent spread of asbestos contamination shall be established including, as applicable, isolation barriers, tenting, local ventilation, scaffolding, clean rooms, shower rooms, equipment

rooms, vacuums, special protective clothing, respiratory protection equipment, sampling equipment, etc.

6.32.5 Air sampling, including pre-work, post-removal, personnel, excursion, etc. monitoring shall be performed in accordance with the applicable regulations and standards.

6.32.6 Asbestos-bearing material that is adjacent to the work area and will not be removed shall be properly sealed and labelled, as applicable.

6.32.7 Construction methods shall be established to minimize the potential for personnel injury. All work shall be performed in accordance with OSHA requirements.

6.32.8 A Pre-ALARA Review Checklist will be completed prior to implementation of this work.

6.33.0 6.33.1 Environmental uglification Re uirements A 10CFR50.49 (EQ) Applicability checklist has been completed and is included as Attachment 1.

6.34.0 Uni ue Re uirements Not Applicable.

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6.34. 1 Design Criteria DC-10034A EWR 10034 Page 89 Revision C

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SECTION 7.0 STEAM GENERATOR SECONDARY MANWAY ACCESS PLATFO S

7.1.0

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7.1.1 7.1.1.1 7.1.1.2 Summar Descri tion of the Desi n Summary In order to facilitate the removal and reinstallation of the steam generators, portions of the existing seconday manway access platforms will be temporarily removed and reinstalled.

In order to facilitate access to the replacement steam generator secondary

manways, permanent secondary manway access platforms will be designed and installed on the steam domes for both the "A" and "B" steam generators.

The platforms will consist of inner and outer circular structural elements which will be used to carry platform loads'hese beams will be supported by posts mounted directly to support lugs on the top of the new steam generators.

The outer circular beam will be supported to the steam dome forging at 4 lug locations and the inner circular beam will be supported at 2 lug locations.

These lugs will be fabricated and installed by the steam generator equipment supplier as part of the manufacturing process.

7. 1.1.3 In addition, for the A steam generator, a walkway between the existing platform at elevation 310'-0" and the new platform will be provided.

RGGE:

Aren't modifications also re ired to the existin B

latform to allow ersonnel to ste off?

7.1.1.4 The temporary removal and re-installation of the snubber hydraulic tubing attached to the existing secondary manway access platforms is addressed in Section 3.0 of this Design Criteria.

Steam generator level and main steam flow instrumentation is addressed in Design Criteria Document DC-10034C.

7.1.2 7.1.2.1 Function The function of the permanent secondary manway platforms is to provide safe access to the secondary manway and a safe and stable working or staging surface.

Additionally, the platform will be designed to integrate necessary removal fixtures and storage areas for the manway covers if required.

7.1.3 7.1.3.1 Performance Requirements None.

Control 7.1.4 Design Criteria DC-10034A EWR 10034 Page 90 Revision C

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7. 1.4. 1 7.1.5 Not Applicable.

Modes of Operation 7.1.5.1 The access platforms will be capable of, and ar' envisioned to, remain attached to the replacement steam generators during all modes of plant operation.

7.1.5.2 7.2.0 7.2.1 RG&E:

Need to discuss mode and or other safet restrictions on installation.

Referenced Documents Rochester Gas 6 Electric, Ginna Station, Quality Assurance Manual.

7.2.2 Equipment Speci fication for Replacement Steam Generators ME-309, Revision 2, 9/1/93.

7.2.3 7.2.4 7.2.5 7.2.5.1

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7.2.6 7.2.6.1 Nuclear Regulatory Commission, NUREG-0800 Standard Review Plan, Section 3.8.4, "Other Seismic Category I Structures",

Revision 1, July 1981.

Nuclear Regulatory Commission, General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR 50.

ly American Institute of Steel Construction (AISC)

Manual of Steel Construction, 9th edition.

American Welding Society (AWS)

AWS D1.1-90, Structural Welding Code - Steel.

7.2.7 Occupational Safety'nd Health Administration 7.2.7.1 General Industry Safety and Health Regulations, Parts 1910.23, 1910.24, 1910.27.

7.2.8 7.2.8.1 7.3.0 7.3.1 RGEE Specifications ME 309, Rev 2.

RGEE to rovide com lete title.

Seismic Cate or The new secondary manway access platforms perform no safety related function, but will be designed and installed as Seismic Category I to ensure that they will not fail and damage safety related components or systems.

RG&E: If latform is non-safet related then seismic classification should be II I not Cate or I.

Su est this be reworded to read as follows:

Design Criteria DC-10034A EWR 10034 Page 91 Revision C

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7.3.1 The new steam generators are classified as safety-related, seismic Category I components.

7.3.1 The existing steam generator seconday manway access platforms are classified as safety-

related, seismic Category I structures.

7.3.2 The new steam generator secondary manway access platforms are classified as safety significant, seismic Category II/I structures and shall be designed and analyzed to ensure that they will not fail and adversely impact safety-related equipment during a seismic event.

7.4.0 ualit Grou Code Class for fluid s stems onl 7.4.1 Not Applicable.

RGSE:

Should the quality group/code class of the steam generators and SGLI be identified since interfaces exist?

7.5.0 7.5.1 7.6.0 7.6.1 Electrical S stem Safet Classification Not Applicable.

A Pro ram A licabilit Per section 3.1.4.32 of Appendix A to.the. Quality Assurance Manual, structures whose failure could prevent safety related equipment from performing its safety functions shall be classified as safety significant.

Therefore, the Quality Assurance Program Requirements as specified in section 3.1.4 shall apply to all activities associated with this installation.

RGRE:

Is a reference to the Bechtel APP needed if Bechtel is oin to install the latforms?

7.7.0 7.7.1 Codes Standards and Re lato Re irements Nuclear Regulatory Commission, NUREG-0800 Standard Review Plan, Section 3.8.4, "Other Seismic Category I Structures",

Revision 1, July 1981.

7.7.2 Nuclear Regulatory Commission, General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR 50.

RGGE:

Wh are the GDC referenced?

Which one is a

licable?

What direct im act do the GDC have on this desicen?

Design Criteria DC-10034A EWR 10034 Page 92 Revision C

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American Institute of Steel Construction (AISC)

Manual of Steel Construction, 9th edition.

7.7.4 American Welding Society (AWS) 7.7.4.1 7.7.5 7.7.5.1 AWS D1.1-90, Structural Welding Code

- Steel.

RGEE:

Wh is 1990 version referenced?

Current edition is 1994.

Occupational Safety and Health Administration General Industry Safety and Health Regulations, Parts 1910.23, 1910.24, 1910..27 RG&E:

What are the effective dates?

RGEE:

Is reference to the UFSAR Gilbert FRS or BWNS seismic data other EWRs needed?

7.8.0 7.8.1 Desi n Conditions Not Applicable.

RG&E:

The im act of ermanentl addin steel inside the containment should be evaluated for an adverse effects on containment s ra covera e

do latforms block assumed s ra

. covera e?

containment heat sink anal ses h dro en eneration anal ses if zinc added containment sum clo in uglification of coatin s

hea

.load dro anal ses if >2000lbs and if latforms ma be ri ed on off with fuel in vessel Does a

minimum seismic "shake-s ace" between the new and existin latforms need to be identified?

7.9.0 7.9.1 Load Conditions The following loads are to be considered during analysis:

Dead Load:

Weight of structural steel,

grating, spokes, handrail, etc.

Live Load:

As specified. by OSHA sections 1910.23 and 1910.27 for handrails and ladders respectively.

RGsE:

Where is wei ht of manwa covers considered?

OBE Loads:

Loads caused by an operating basis earthquake.

SSE Loads:

Loads caused by a safe shutdown earthquake.

Design Criteria DC-10034A EWR 10034 Page 93 Revision C

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Thermal Stress:

Stress due to the heat transfer caused by the direct contact between the lifting lugs on the steam generator steam dome and the platform.

RG&E:

To be consistent with the other sections of this DCD re ared b

Bechtel should load combinations and acce tance criteria be identified here?

7.10.0 7.10.1 Environmental Conditions The new steam generator seconday manway access platforms shall be designed for the following environmental conditiOns in accordance with RG&E to identif reference:

Normal Accident Ambient Temperature

('F) 41 - 120 300 Ambient Pressure (psig) 0 3

60 Relative Humidity (4) 21 100 100 RG&E:

Should the environmental conditions identified here be the sam as those list'ed in Section 7.10.1 er EWR 5275?

7.11.0 Interface Re irements 7.11.1 The installation and used of these platforms shall not restrict access to existing plant systems or equipment.

RG&E:

Should the interface with the new steam enerators and the existin manwa latforms be identified?

7.12.0 7.12.1 Material Re irements All materials used in building the platforms shall be purchased, stored and handled in accordance with Ginna Station Quality Assurance Program.

7.13.0 7.13.1 RG&E To be consistent with the other sections of this DCD re ared b

Bechtel should s ecific material t

es coatin s

weldin materials weld rocedures etc.

be identified?

Mechanical Re irements Not Applicable.

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7.14.0 Design Criteria DC-10034A EWR 10034 Page 94 Structural Re uirements Revision C

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7. 14. 1 Allowable stresses for structural steel shall be in accordance with the AISC Manual of Steel Construction, 9th edition.

7.14.2 The acceptance criteria for both steel and concrete shall be as specified in NRC Standard Review Plan Section 3.8.4, Paragraph II.5.

RGEE:

What concrete is involved?

7. 14. 3 The additional mass attached to the steam generator vessel shall be included in the structural analysis calculation performed by the steam generator supplier and design vessel stresses shall be within allowable stresses in accordance with RGEE Specification ME 309.

RGaE:

The followin pints should be considered:

(1)

Design and acceptance criteria for modifications to the existing platforms should be identified.

(3)

Design and acceptance criteria for the steam generator supports should be identified for the evaluation of the increased weight.

(Part of the BWNS evaluation?)

7. 15.0 7.15.1 7.16.0 H draulic Re irements Not Applicable.

Chemist Re irements 7.16.1 7.17.0 7.17.1 7.18.0 7.18.1 7.19.0 7.19.1 7.20.0 7.20.1 7.21.0 7.21.1 Not Applicable.

Electrical Re irements Not Applicable.

La out and Arran ement Re irements Not Applicable:

0 erational Re irements Not Applicable.

Instrumentation 0 Control Re irements Not Applicable.

Access a Administrative Control Re irements Not Applicable.

7.22.0 Design Criteria DC-10034A EWR 10034 Revision C

Page 95 Redundanc Diversit 6

Se aration Re uirements

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7.22.1 7.23.0 Not Applicable.

Failure Effects Re uirements 7.23.1 7.24.0 7.24.1 7.25.0 The steam generator secondary manway access platforms shall maintain their structural integrity when they are subjected to the design basis seismic event.

Therefore, there will be no no adverse effects on safety-related systems or equipment.

Test Re uirements Not Applicable.

Accessibilit Maintenance Re air R Xnservice Zns ection Re uirements 7.25.1 7.26.0 7.26.1 7.27.0 7.27.1 7.28.0 7.28.1 Not Applicable.

Personne1 Re irements Not Applicable.

Trans ortation Re irements Not Applicable.

Fire Protection Re irements This installation will be reviewed per Engineering Procedure QE-326 to ensure compliance with the applicable provisions of 10 CFR 50 Appendix R and the Facility Operating License.

RG&E:

Should wordin similar to that in luded in Section 4.28 be ado ted here Handlin Re irements 7.29.0 7.29.1 7.30.0 7.30.1 7.31.0 7.31.1 7.32.0

7. 32. 1 Not Applicable..

Public Safet Re irements Not Applicable.

A licabilit Not Applicable.

Personnel Safet Re irements reg Design Criteria Revision C

The design and installation of the platforms shall consider all applicable requirements of the OSHA ulations.

DC-10034A EWR 10034 Page 96 Date 09 26 94

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rt 7 '2.2 Radiation protection measures shall be in place during the performance of the work.

Good housekeeping practice shall be maintained.

All work shall comply with the Rochester Gas

& Electric Accident Prevention Nanual.

7.32.3 7.33.0 7.33.1 7.34.0 7.34.1 A Pre-ALARA Review Checklist will be completed prior to implementation of this work.

Environmental uglification Re uirements A 10CFR50.49 (EQ) Applicability checklist has been completed and is included as Attachment 1.

Uni ue Re uirements Not Applicable.

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t Design Criteria DC-10034A EWR 10034 Page 97 Revision C

Date 09 26 94

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