ML21123A226
| ML21123A226 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 10/29/2020 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WBL-20-047 | |
| Download: ML21123A226 (639) | |
Text
{{#Wiki_filter:WBN 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8.1 Concrete Shield Building The Shield Building is a Category I structure in its entirety and is designed to remain functional in the event of a Safe Shutdown Earthquake (SSE) or a tornado. The Shield Building is designed as described in Sections 3.8.1.1 through 3.8.1.7. The evaluation and modification of the Shield Building reinforced concrete structure are optionally done using the ultimate strength design method in accordance with the codes, load definitions and load combinations specified in Appendix 3.8E. 3.8.1.1 Description of the Shield Building The Shield Building, shown in Figure 3.8.1-1 through 3.8.1-7, is a reinforced concrete structure surrounding the steel containment structure and is designed to provide the following: radiation shielding from accident conditions, radiation shielding from parts of the Reactor Coolant System (RCS) during operation, and protection of the steel containment vessel from adverse atmospheric conditions and external missiles propelled by tornado winds. The Shield Building is a reinforced concrete cylinder supported by a circular base slab and covered at the top with a spherical dome. It is located adjacent to the concrete Auxiliary and Valve Room Buildings and is physically separated from them by a 1-inch fiberglass-filled expansion joint. There is a polyvinyl chloride seal placed in formed grooves on the face of the Shield Building where it abuts the Auxiliary Building, thus providing water tightness between the two buildings up to grade level of Elevation 728.0. The seal is embedded in the groove with epoxy adhesive mortar. The Shield Building is maintained watertight to Elevation 742.0. A sectional view through the Shield Building is shown in Figure 3.8.1-1. Only the base slab resists the LOCA pressure load which is transmitted to it through a steel plate liner anchored to its top face. For further discussion of the base slab see Section 3.8.5. The cylinder wall is approximately 150 feet in height from the top of the base slab to the spring line of the dome. It has an inside diameter of 125 feet 1 inch and a thickness of 3 feet. Conventional steel reinforcing bars were used throughout the structure and were placed in a horizontal and vertical pattern in each face of the cylinder wall. The area of reinforcement in each direction of each face is not less than 0.0015 times the gross concrete area. The effects of penetrations through the wall were considered. Penetrations, 12 inches or less in diameter, do not significantly disturb the reinforcing pattern in the wall. Therefore, no special reinforcing considerations were made at these areas. For penetrations larger than 12 inches, reinforcing is terminated at the opening. Supplemental reinforcing is added, both vertically and horizontally, to replace the reinforcing, terminated at rectangular penetrations larger than 12 inches and circular penetrations larger than 24 inches. The amount of supplemental reinforcing added is equal to or greater than the amount of 3.8-1
WBN reinforcing removed and is placed adjacent to the penetration. In addition, rectangular penetrations in the wall have diagonal reinforcing across the corners. Reinforcing bars were lap spliced in accordance with ACI 318-71 code requirements for strength design or have been cad-welded. Reinforcing steel bars in the dome were arranged in a radial and circumferential pattern. A ring tension beam is provided to resist the outward thrust from the dome roof. The tensile force in the ring beam is resisted by 24 No. 11 reinforcing bars. These bars are spliced with mechanical splices that are uniformly staggered at least 6 feet on center around the circumference of the ring beam. Therefore, at any cross section in a length of 6 feet, only three bars are spliced out of the total of 24 bars, and not more than two of these are in any one layer. That is, at any section, 21 bars are continuous and unspliced. These continuous, unspliced bars alone will carry the imposed load with only a 15 percent increase in stress. Stirrups enclosing the main reinforcement are spaced on 15-inch centers. To facilitate removal of the old steam generators (OSGs) and installation of the replacement steam generators (RSGs) during the Unit 1 steam generator replacement (SGR), two construction opening were cut in the concrete shield building dome. These openings were restored by splicing new reinforcing bar to the existing reinforcing bar using Bar-Lock couples, Cadwelds, and/or welding and pouring new concrete to close the openings. 3.8.1.1.1 Equipment Hatch Doors and Sleeves As shown in Figure 3.8.1-8, a double-leaf equipment door installed in a sleeve is provided for each Reactor Building. The steel sleeve forms an access through the Shield Building wall to the equipment hatch in the Containment Vessel. Each sleeve extends from inside the Shield Building to the shielded passageway leading to the Auxiliary Building floor Elevation 757.0. Each door is of the hinged, double-leaf, marine type with seals for providing an airtight closure between the annulus surrounding the steel containment vessel and the inside of the Auxiliary Building. A door will normally be opened only when the reactor is in the shutdown, depressurized condition such that secondary containment is not required. The sleeves, embedded in the Shield Building walls, are of welded steel construction, rectangular in cross section, with corners fabricated to a radius. They form clear passageways 20 feet wide and 17 feet-8 inches high through the concrete walls of the Shield Buildings. Floors in the sleeves are at Elevation 756.63 coinciding with the Elevation of the operating floors in the Reactor Buildings. The doors are hinged to the sleeves on the end toward the outside of the Shield Building wall and are of welded construction consisting of structural shapes with a steel skin plate. 3.8-2
WBN Sealing of a door when closed is by means of solid, molded rubber seals mounted on the door. The seals contact the edge of the sleeve at the top and sides, a removable seal bar at the floor level, and a sealing bar at the meeting line of the two leaves. Penetrations through the doors are sealed with solid rubber O-ring type seals. The doors are opened and closed manually. Latching of the doors in the closed position is accomplished by hand-lever operated dogs acting on wedge surfaces around the perimeter and meeting edges of the door leaves. The doors are part of the airtight closure between the annulus surrounding the Containment Vessel and the inside of the Auxiliary Building. These doors are to remain closed during unit operation and will only be opened during unit shutdown. The door and sleeves will maintain their structural integrity and remain operational after being subjected to the environmental or accident conditions listed in Section 3.8.1.4. 3.8.1.2 Applicable Codes, Standards, and Specifications The structural design of the reinforced concrete Shield Building is in compliance with the proposed ACI-ASME (ACI-359) Code for Concrete Reactor Vessels and Containment, Article CC-3000, as issued for trial use, April 1973, for the loading combinations defined in Table 3.8.1-1. Allowable stresses are based on this code with the exception of allowable tangential shear stresses in walls where the ACI 318-71 code is used. Detailing of reinforcing around opening of circular walls is based on the ACI Chimney Code (ACI 307-69), Sections 4.4.4 through 4.4.7. All reinforcing steel conforms to the requirements of ASTM Designation A615-72, Grade 60. Unless otherwise indicated in the UFSAR, the design and construction of the Shield Building is based upon the appropriate sections of the following codes, standards, and specifications. Modifications to these codes, standards, and specifications are made where necessary to meet the specific requirements of the structures. Where date of edition, copyright, or addendum is specified, earlier versions of the listed documents were not used. In some instances, later revisions of the listed documents were used where design safety was not compromised.
- 1. American Concrete Institute (ACI)
ACI 214-77 Recommended Practice for Evaluation of Strength Test Results of Concrete ACI 318-71 Building Code Requirements for Reinforced Concrete ACI 359 Code for Concrete Reactor Vessels and Containments, (Proposed ACI-ASME Code ACI-359 (Article CC-3000) As issued for trial use April, 1973) 3.8-3
WBN ACI 347-68 Recommended Practice for Concrete Formwork ACI 305-72 Recommended Practice for Hot Weather Concreting ACI 211.1-70 Recommended Practice for Selecting Proportions for Normal Weight Concrete ACI 307-69 Specification For the Design and Construction of Reinforced Concrete Chimneys
- 2. American Institute of Steel Construction (AISC)
'Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings,'
adopted February 12, 1969, except welded construction is in accordance with Item 4 below.
- 3. American Society for Testing and Materials (ASTM), 1975 Annual Book of ASTM Standards. Specific standards are identified in Section 3.8.1.6.
- 4. American Welding Society (AWS)
Structural Welding Code, AWS D1.1-72 with Revisions 1-73 and 2-74 except later editions may be used for prequalified joint details, base materials, and qualification of welding procedures and welders. Visual inspection of structural welds will meet the minimum requirements of Nuclear Construction Issues Group documents NCIG-01 and NCIG-02 as specified on the design drawings or other engineering design output. See Item 12 below.
'Recommended Practice for Welding Reinforcing Steel, Metal Inserts, and Connections in Reinforced Concrete Connections,' AWS D12.1-61.
- 5. Uniform Building Code, International Conference of Building Officials, Los Angeles, 1970 edition.
- 6. Southern Standard Building Code, 1969 edition, 1971 Rev.
- 7. 'Nuclear Reactors and Earthquakes,' USAEC Report TID-7024, August 1963.
3.8-4
WBN
- 8. American Society of Civil Engineers Transactions, Volume 126, Part II, Paper No. 3269,
'Wind Forces on Structures,' 1961.
- 9. Code of Federal Regulations Title 29, Chapter XVII, "Occupational Safety and Health Standards," Part 1910.
- 10. NRC Regulatory Guides; RG 1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures RG 1.12 Instrumentation for Earthquakes RG 1.15 Testing of Reinforcing Bars for Category I Concrete Structures RG 1.31 Control of Ferrite Content in Stainless Steel Weld Metal RG 1.55 Concrete Placement in Category I Structures.
- 11. Nuclear Construction Issues Group (NCIG)
NCIG-01, Revision 2 - Visual Welding Acceptance Criteria (VWAC) for Structural Welding NCIG-02, Revision 0 - Sampling Plan for Visual Reinspection of Welds The referenced NCIG documents may be used after June 26, 1985, for weldments that were designed and fabricated to the requirements of AISC/AWS. NCIG-02, Revision 0, was used as the original basis for the Department of Energy (DOE) Weld Evaluation Project (WEP) EG&G Idaho, Incorporated, statistical assessment of TVA performed welding at WBNP. Any further sampling reinspections of structural welds subsequent to issuance of NCIG-02, Revision 2, are performed in accordance with NCIG-02, Revision 2 requirements. The applicability of the NCIG documents is specified in controlled design output documents such as drawings and construction specifications. Inspectors performing visual weld examination to the criteria of NCIG-01 are trained in the subject criteria.
- 12. TVA Reports CEB 86-12 Study of Long Term Concrete Strength at Sequoyah and Watts Bar Nuclear Plants CEB 86-19-C Concrete Quality Evaluation 3.8-5
WBN 3.8.1.3 Loads and Loading Combinations The Shield Building dome and cylinder wall are subjected to the following loads. Design loading combinations utilized to examine the effects of localized areas are shown in Tables 3.8.1-1 and 3.8.1-2. Dead Load This includes weight of the concrete structure plus any other permanent load contributing to stress, such as equipment, piping, and cable trays suspended from the structures. Earth Pressure The static soil pressure was computed using Earth Pressure Standards from TVA's General Standards which incorporate Coulomb's "wedge of pressure" theory. Standard soil properties for fine grained rolled fill are as follows: Angle of internal friction = 32 degrees Angle of friction between soil and building = 16 degrees Dry weight = 120 lb/cu ft Buoyant weight = 65 lb/cu ft Due to adjacent structures the soil does not completely surround the Shield Building but lies in a 185-degree segment around it. The soil was backfilled to a height of 31 feet above the base slab. A surcharge of 200 psf was used. Hydrostatic Pressure Uplift forces and lateral static pressure were computed using the full hydrostatic head measured from the water surface. Water surface elevations from the probable maximum flood (Section 2.4) were used in determining hydrostatic heads. Due to water seals between the Shield Building and adjacent structures, the lateral hydrostatic pressure was applied only to one-half of the circumference for the drawn down ground water table. For the probable maximum flood the adjacent structures are allowed to flood and lateral hydrostatic pressure was applied around the full circumference. 3.8-6
WBN Loss-of-Coolant Accident (LOCA) In addition to the reactions of the containment vessel and interior concrete due to the LOCA pressure transients, the LOCA produced uplift forces on the steam generator or reactor coolant pump anchors in the base slab. The LOCA also increased the temperature in the annulus space between the Containment Vessel and the Shield Building. This produced a nonlinear temperature gradient across the cylinder wall and dome. A typical gradient in shown in Figure 3.8.1-9. Normal Temperature Gradient The temperature gradient for normal plant operation was considered as uniformly varying through the section. The maximum temperature gradient occurs just above grade when the plant is in operation and a minimum ambient temperature exists. The normal temperature difference across the wall varies from a minimum of 35°F below grade to the maximum of 85°F as shown on Figure 3.8.1-9. Operational Basis Earthquake (OBE) The plant was designed to remain operational under the OBE. The OBE has a maximum acceleration of 0.09g horizontally and 0.06g vertically. In addition to the maximum values of the structural response in terms of displacement, acceleration, shear, moment, torque and axial force, the soil pressure and hydrostatic pressures were increased due to seismic motions. The static soil pressure was increased 23% for a dry fill and 11% for a saturated fill. This incremental increase was a triangle of pressure with the apex at the rock surface and the maximum ordinate at the ground surface. The hydrostatic pressure of the water within the fill was increased by 11%. This incremental increase was a triangle of pressure with the apex at the water surface and maximum ordinate at the rock surface. The magnitude of these increases were determined by shaking table experiments performed for another TVA project. The reaction from earthquake motion on the compressed expansion joint material separating the adjacent Auxiliary and Valve Room Buildings was also taken into consideration. Safe Shutdown Earthquake (SSE) The plant was designed to have the capability for safe shutdown for the SSE (maximum acceleration of 0.18g horizontally and 0.12g vertically). The incremental pressure increase for soil and hydrostatic pressure was twice that for the OBE. Live Load Live load includes non pipe hanger loads, plus any other permanent load such as crane loads, etc. Snow load of 20 psf was considered in the design live load. 3.8-7
WBN Tornado The tornado was assumed to have an "eye" whose pressure is 3 psi below ambient, a "funnel" having a rotational velocity of 300 mph, and a translational speed of 60 mph. The Shield Building was designed for wind loads corresponding to 360 mph and a maximum internal pressure of 3 psi. Maximum wind velocity and maximum internal pressure loading do not coincide as shown by Figure 3.3-1. The ultimate capacity of the structure in flexure or shear is not exceeded under the combined pressure and wind velocity loadings of Figure 3.3-1. The adjacent structures disturb the air flow around the Shield Building. The only method to determine the actual pressure distribution on the structure is by a model test. In lieu of model test, several cases of extreme pressure distributions were analyzed in an attempt to bracket the actual stresses. The normal maximum wind loading was based on Figure 1(b), from ASCE Paper 3269, "Wind Forces on Structures." Tornado missiles are described in Section 3.5. Construction Loads - Historical Information The dome was poured in two lifts. The first lift is a 9-inch pour supported by temporary shoring bearing on the Containment Vessel. The first lift was designed to support the wet concrete dead load of the second lift plus a construction load of 50 psf. 3.8.1.4 Design and Analysis Procedures Base Slab The base slab is discussed in Section 3.8.5. Cylinder Wall and Dome The stiffness of the cylinder wall was small in comparison to that of the base slab and the cylinder wall was assumed fixed at the base. The height of the wall was such that the effect of discontinuity at one end was negligible when considering discontinuity at the other end. For symmetrical loadings, the edge forces at the point of discontinuity were determined by writing the equations of the primary system and the equation of compatibility. The discontinuity stresses from the edge forces were superimposed on the membrane stresses. The above analysis was checked by two independent computer analyses ("Axisymmetric Finite Element Analysis, AMG032" and GENSHL 2). Unsymmetrical loadings, such as wind, were analyzed by using computer code, GENSHL. These loads were approximated through a Fourier series. 3.8-8
WBN Creep and Shrinkage Effects Creep was not considered in the design of the Shield Building. Sustained loads are essentially the dead weight loads of the structure itself with subsequent stress levels too low to influence creep deformations to any significant degree particularly since these deformations do not cause differential settlements in the structure. Shrinkage effects are considered in the design of all structures by estimating the temperature change from peak hydration temperatures to final operating temperature conditions. In addition drying shrinkage effects are considered in all members which have an average drying path of less than 15 inches. The methods used to consider these effects are explained in an ACI Committee 207 Report 70-45, "Effect of Restraint, Volume Change, and Reinforcement on Cracking of Massive Concrete" published in July 1973. The effects of base restraint on the cracking of a circular structure is essentially the same as the effects on a wall of equal thickness whose length is equal to the outside diameter of the circular structure. The Shield Building was not only designed to restrict shrinkage cracking, thus holding the cracks to a minimum acceptable size, but was also waterproofed on the exterior surface below grade to eliminate possible seepage. The portion above grade is essentially out of the restraint zone and will therefore be relatively free from shrinkage cracking. Tangential Shear The tangential shears induced by earthquake and wind forces were assumed to vary from zero over a thickness of wall located at the extremes of a diameter parallel to the line of action of the shearing force to a maximum on a wall thickness located at the extremes of a diameter normal to the line of action of the shearing force. Distribution was assumed proportional to the cosine of the polar angle measured from the diameter normal to the line of action of the shear force with a maximum allowable shear stress in the concrete limited to 247 psi according to special provisions for shear in walls in the ACI 318-71 code. Seismic See Section 3.7 for a detailed description of the seismic analysis. 3.8-9
WBN Equipment Hatch Doors and Sleeves For the closed position, the structural members of the door leaves were designed as simple beams under uniformly distributed loading with the end reactions carried by the sleeve. Loads at the dogging wedges were carried to the sleeve as concentrated loads. For the open position, the door leaves were treated as cantilever structures, and the hinge members and sleeve were designed for the resulting concentrated loads. Design of the doors and sleeves was by TVA without the use of a computer program. Under normal operating conditions, air pressure equal to 5 inches of water is exerted on the Auxiliary Building side of the doors. Under accident or tornado conditions, the doors are subjected to air pressure. Environmental and accident conditions which were considered in the design of the doors and sleeves are as follows:
- 1. The OBE and the SSE with accelerations as hereinafter defined.
- 2. An inadvertent release of the cooling sprays in the Containment Vessel will cause a pressure drop within the annulus surrounding it and result in an air pressure load of 2 psi on the Auxiliary Building side of the doors and sleeves. Duration of this condition will be for a few hours maximum.
- 3. A tornado condition which causes a pressure drop within the Auxiliary Building will result in a pressure of 3 psi on the annulus side of the doors. Duration will be for 3 seconds.
- 4. A LOCA accident in the Containment Vessel which will result in a pressure equal to 3/4 inch of water on the Auxiliary Building side of the doors. A partial vacuum is created in the annulus by vacuum pumps, and this condition may exist for a period of several months.
Earthquake accelerations used in design of the doors and sleeves were determined by dynamic analysis of the supporting structure of the Shield Building. Accelerations at the centerline of the equipment hatch for the OBE are as follows: Lateral (north-south) 0.16g Lateral (east-west) 0.16g Vertical 0.12g Accelerations at the centerline of equipment hatch for a SSE are as follows: Lateral (north-south) 0.36g Lateral (east-west) 0.36g Vertical 0.23g 3.8-10
WBN These accelerations were used as static loads for determining component and member sizes. After establishing the component and member sizes, a dynamic analysis, using appropriate response spectrum, was made of each sleeve and its doors to determine that allowable stresses had not been exceeded. 3.8.1.5 Structural Acceptance Criteria Controlling Conditions - Shield Building Structure The SSE in combination with a LOCA (load combination 8) produced the largest overturning moment. For this combination, the percent of the base slab in compression was 51% and the factor of safety for overturning was 1.74. The uplift on the equipment from the LOCA combined with the SSE controlled the design of the base slab. Minimum steel requirements of 0.65 square inches per foot (minimum steel ratio of 0.0015 in each face and in both vertical and horizontal directions) controlled the inside face vertical steel requirements throughout the shell and the inside face horizontal steel requirements above grade. The SSE in load combination 8 controlled the design of the outside face vertical reinforcement at the base of the cylinder wall. Due to earth and hydrostatic pressure, outside face horizontal reinforcement requirements were greatest 16 feet above the base of the cylinder wall at Elevation 713.0. The construction loading controlled the reinforcement design in the dome and the upper portion of the cylinder wall. The SSE produced a maximum tangential shear stress at the base of the wall of 189.7 psi which was 76.8% of the allowable. The effects of repeated reactor shutdowns and startups during the plant's life will not degrade the above margins of safety because the Shield Building is minimally affected by these operations. The only effects from normal operations are from interior temperature changes which are insignificant compared to normal exterior temperature variations. Equipment Hatch Doors and Sleeves Allowable stresses for load combinations used for the various parts are given in Table 3.8.1-2. For normal load conditions, the allowable stresses provide safety factors of 1.67 (Fy/0.6 Fy) to 1 on yield for structural parts and 5 to 1 on ultimate for mechanical parts. For a limiting condition such as a Safe Shutdown Earthquake (SSE), stresses do not exceed 0.9 yield. 3.8-11
WBN 3.8.1.6 Materials, Quality Control and Special Construction Techniques General The principal materials used in the construction of the Shield Building base slab, wall, and dome were concrete and reinforcing steel. Steel is used for the structural parts of the equipment hatch doors and sleeves with rubber used for the seals. 3.8.1.6.1 Materials Concrete Cement conformed to ASTM Specification C150-72 Type I. The guaranteed 28-day mortar strength was 5025 psi with a guaranteed standard deviation of 395 psi and a guaranteed maximum tricalcium aluminate content of 9.5%. Aggregates conformed to ASTM Specification C-33-71a and were manufactured of crushed limestone. Water for mixing concrete and also for washing the aggregates and curing concrete was tested prior to use in accordance with Corps of Engineers test method CRD-C400. The fly ash used at Watts Bar is in general accordance with the ASTM C618-73, except for the loss of ignition and fineness of pozolanic index parameters. TVA specific requirements for loss of ignition are more restrictive while the fineness pozolanic index is less restrictive than the ASTM requirements. (See Section 3.8.3.2.1.a for more details). Sampling and testing was performed in accordance with ASTM C 311. Air-entraining admixtures conformed to ASTM Specification C-260-69. Water-reducing agent used for concrete mixtures containing fly ash was selected based on demonstrated achievement of TVA specified concrete strength of a control mix by actual testing. Reinforcing Steel Reinforcing steel conformed to ASTM Designation A615-72, Grade 60. For the Unit 1 steam generator replacement, reinforcing steel used in the restoration of the shield building construction openings conformed to ASTM A615, Grade 60. Bar-Lock Couplers During the Unit 1 steam generator replacement, Bar-Lock couplers were used to splice the new reinforcing bar to the existing reinforcing bar during the restoration of the shield building construction openings. Bar-Lock couplers are manufactured from seamless hot-rolled steel tube conforming to ASTM A-519 specification, with minimum tensile strength exceeding 100,000 psi. 3.8-12
WBN Equipment Hatch Sleeves and Doors The structural parts of the sleeves and doors are fabricated from ASTM A36 steel. 3.8.1.6.2 Quality Control Concrete Concrete was produced in a central batch and mixing plant until 1977, and central batch and transit mix after 1977. A materials engineering unit was specifically responsible for control, documentation, and daily review of test data. Aggregate gradation and deleterious material was checked daily. All coarse aggregate was rinsed and resized. The gradation of the fine aggregate and the amount finer than the No. 200 sieve conformed to specifications. The other concrete material was also subject to periodic tests (see Section 3.8.3.2). The specified strength of the concrete was 4000 psi at 28 days. Some concrete did not meet specification requirements. This was evaluated and documented in the Report CEB-86-19-C "Concrete Quality Evaluation." The results have been documented in affected calculation packages and drawings. A testing program conducted at the site compared strengths of cylinders and concrete from 3-foot-thick wall sections subjected to exterior exposures. The results of this test program are documented in TVA report CEB 86-12, "Study of Long-Term Concrete Strength at Sequoyah and Watts Bar Nuclear Plants." These tests demonstrated the long term compressive strength gain with age which have occurred. The strength gain and age was generally 2600 psi beyond 28 days and 1300 psi beyond 90 days. During the Unit 1 steam generator replacement, concrete used for the restoration of the shield building dome construction openings was provided in accordance with Specification 24900-C-321. The concrete was designed to achieve a minimum strength of 4000 psi at seven days. Reinforcing Steel Testing of reinforcing steel conformed to Regulatory Guide 1.15. Cadweld splices conformed to Regulatory Guide 1.10. Bar-Lock Couplers Manufacturing processes and procedures for the Bar-Lock couplers used in the restoration of the Shield Building openings flowing installation of the replacement steam generators complied with the applicable provisions of ANSI/ASME N45.2. Qualification testing of the Bar-Lock couplers conformed to ASME Code, Section III, Division 2, CC-4333.2, Splice System Qualification Requirements. 3.8-13
WBN Equipment Hatch Doors and Sleeves Design by TVA and erection by TVA were in accordance with TVA's quality assurance program. Design and fabrication by the contractor were in accordance with the contractor's quality assurance program which was reviewed and approved by TVA's design engineers. The contractor's quality assurance program covers the criteria in Appendix B of 10 CFR 50. Fabrication procedures such as welding and nondestructive testing were included in appendices to the contractor's quality assurance program. ASTM standards were used for the material specifications and certified mill test reports were provided by the contractor for materials used for load carrying members. Material used for seals including O-rings, was certified by a rubber technologist as being capable of withstanding the radiation and temperature conditions existing during a LOCA accident. This certification is based on testing and evaluation of seal materials performed under contract for TVA by Presray Corporation. 3.8.1.6.3 Construction Techniques - Historical Information The walls of the Shield Building from the base slab to the bottom of the ring beam were constructed using conventional forms. The concrete pouring was performed in two stages to facilitate other construction work in the building. The first stage consisted of concrete pours to Elevation 762.0 and the second stage consisted of the remaining height of wall. Concrete temperatures were monitored throughout for a minimum period of 3 days during cold weather to assure cold weather protection requirements. The dome roof was placed in two lifts with each lift divided into three basic rings and each ring divided into radial segments. The Steel Containment Vessel (SCV) is designed to support the form work for the first 9-inch-thick lift and the first lift is then designed to support the remaining 15-inch lift with the form work removed. Delays are specified between adjacent lift pours in order to minimize the effects of initial volume changes. The second lift was not placed until the first lift had attained its specified strength. The base slab, ring beam, and parapet wall were constructed using conventional methods. 3.8.1.7 Testing and Inservice Surveillance Requirements Since the Shield Building is not a pressure containment, its wall and dome will not be pressure tested. REFERENCES None 3.8-14
WBN TABLE 3.8.1-1 (SHEET 1 of 2) LOADING COMBINATIONS, LOAD FACTORS AND ALLOWABLE STRESSES FOR THE SHIELD BUILDING CONCRETE EXTERIOR CYLINDRICAL WALL, DOME AND BASE SLAB COMBINATIONS(3) LOADING 1 2 2a 2b 3 4 5 6 7 8 9 D DEAD LOAD 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 10 D EARTH PRESSURE 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 PMF PROBABLE MAXIMUM FLOOD 1.0 To NORMAL OPERATING TEMPERATURE 1.0 1.0 1.0 1.0 1.0 1.0 Ta ACCIDENT TEMPERATURE 1.0 1.0 1.0 1.0 Pa ACCIDENT PRESSURE 1.5 1.25 1.25 1.0 Fegs SAFE SHUTDOWN EARTHQUAKE 1.0 1.0 Fego OPERATIONAL BASIS EARTHQUAKE 1.0 1.0 1.25 W NORMAL WIND 1.0 1.0 1.25 1.0 Wt TORNADO(2) 1.0 L LIVE LOAD 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 CC CONSTRUCTION CONDITION 1.0 Pv NEGATIVE INTERNAL PRESSURE 1.0 1.0 1.0 YjYr PIPE BREAK JET AND REACTION LOAD 1.0 1.0 1.0 fc .45fc' .45fc' .45fc' .45fc' .75fc' .75fc' .75fc' .75fc' .75fc' .75fc' .75fc' ALLOWABLE STRESSES* (1) (1) (1) (1) fs .5 fy .5fy .5 fy .5fy .9 fy .9 fy .9 fy .9 fy .9 fy .9 fy .9 fy
*fc' = SPECIFIED STRENGTH OF CONCRETE fc = ALLOWABLE FLEXURAL CONCRETE STRESS fy = YIELD STRENGTH OF REINFORCING STEEL fs = ALLOWABLE REINFORCING STEEL STRESS FOOTNOTES:
(1) REINFORCING STEEL STRESSES MAY BE INCREASED BY 33% WHEN TEMPERATURE EFFECTS ARE COMBINED PROVIDED THE REQUIRED SECTION IS NOT REDUCED FROM THAT REQUIRED WITHOUT THE TEMPERATURE EFFECTS (2) Wt INCLUDES TORNADO WIND, TORNADO POSITIVE INTERNAL PRESSURE, AND TORNADO GENERATED MISSILES.
WBN TABLE 3.8.1-1 (SHEET 2 of 2) FOOTNOTES (Continued): (3) LOADING COMBINATIONS (COMPARED TO TABLE CC-3200-1 OF ACI-359, 1973)
- 1. Service - Construction
- 2. Service - Normal
- 3. Factored - Extreme
- 4. Factored - Environmental
- 5. Factored - Abnormal
- 6. Factored - Abnormal/Severe Environmental
- 7. Factored - Abnormal/Severe Environmental
- 8. Factored - Abnormal/Extreme Environmental
- 9. Factored - Extra Case The following loads from Table CC-3200-1 of ACI-359, 1973, as issued for trial use, are not applicable to the Shield Building exterior wall and dome.
(F, Pt, Tt, Ro, Ra, Yr, Yj, Ym, Pa, Ta) = 0 The Structural Integrity Test (D + L + Pt + Tt) from the ACI-359, 1973 is not a controlling load case for the base slab.
WBN TABLE 3.8.1-2 (SHEET 1 of 2) SHIELD BUILDING EQUIPMENT HATCH DOORS AND SLEEVE LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Structural No. Load Combinations Allowable Stresses (psi) Tension Compression**** Shear I Dead load plus 2-psi pressure 0.50 Fy 0.47 Fy 0.33 Fy II Dead load plus 3-psi pressure 0.90 Fy 0.90 Fy 0.60 Fy inside III Dead load plus 2-psi pressure 0.60 Fy 0.60 Fy 0.40 Fy outside plus *OBE IV Dead load plus 2-psi pressure 0.90 Fy 0.90 Fy 0.60 Fy outside plus *SSE
**V Dead load plus *OBE 0.60 Fy 0.60 Fy 0.40 Fy **VI Dead load plus *SSE 0.90 Fy 0.90 Fy 0.60 Fy Mechanical No. Load Combinations Allowable Stresses (psi)
Tension & Shear Compression(****)
**I Dead load Ult 2 x Ult 5 15 ***Ia Dead load plus* OBE 0.60 Fy 0.40 Fy ***II Dead load plus *SSE 0.90 Fy 0.60 Fy
WBN TABLE 3.8.1-2 (Sheet 2 of 2) SHIELD BUILDING EQUIPMENT HATCH DOORS AND SLEEVE LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES (Cont'd) III Dead load plus 2-psi pressure outside Ult 2 x Ult 5 15 IV Dead load plus 3-psi pressure inside 0.90 Fy 0.60 Fy V Dead load plus 2-psi pressure outside 0.60 Fy 0.40 Fy plus *OBE VI Dead load plus 2-psi pressure outside 0.90 Fy 0.60 Fy plus *SSE
- Acts in one horizontal direction only at any given time and acts in the vertical and horizontal directions simultaneously.
** Door open. *** For hinges only with doors open.
- The value indicated for allowable compression stress is the maximum value permitted when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls.
Kl 2 r when K1 < Fcr = FY 1 - Cc 2 Cc 2 r or E 2 Kl Fcr = 2 when > Cc Kl r 0 r
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<KBtrY1____lfir. ep 4MtfafAt UNIT 2 NN9Po6i-12_CWORETE PIXtR ptAWTPoG CONCRETE EXTERIOR WALL FOR HATCHED AREAS SEE UNIT I DRAWING 4IN712-I OUTLINE TVA DWG NO. 2-41N712-1 MA COWANION DWG; 2-4IN712-2 ! 3 FIGURE 3.8.1-4(U2)
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- 1 SHOWN ON THIS DRAWING DERIVED FROM REPORT CEB ad-19--V. SEE NOTE 18 8°0 VENT HOLE V PENT HOLE n (TYP, 4 PLCS) 0°(t2°)
SEE NOTE 17
- 18. UNIT 2 ONLY - MINIMUM ACTUAL STRENGTH OF CONCRETE IS 4800 PSI FOR tet LIFTS AND 4100 PSI FOR 2n4 LIFTS.
- 17. FIELD TO FILL HOLES WITH EMACO $56 CI WHEN HOES ARE NO LONER REQUIRED.
UFSAR AMENDMENT 1 WATTS BAR FINAL SAFETY ANALYSIS REPORT REACTOR BUILDING UNIT 2 FOR HATCHED AREAS SEE UNIT 1 AC DWG 41N718-1 CONCRETE DOME fU°OM DRAWINGS, 41W18 _ BILL OF MATERIAL 41NlMW-lW_ MNMETE POUR DRAWING OUTLINE TVA DWG NO. 2-41N718-1 R1 FIGURE 3.8.1-7(U2)
3=0" WALL INSIDE FACE OUTSIDE FACE 0 6 /2 /8 24 30 36 INCHES SH/ELD BUILDING -EL 728 TO EL 745 WATTS BAR NUCLEAR PLANT FINAL SAFETY WATTS BAR NUCLEAR PLANT FINAL SAFETY SHIELD BUILDING ANALYSIS REPORT TEMPERATURE GRADIEiv'T elevation 728-745 Shield FigureBuilding 3.8.1-9 Temperature Gradient Elevation 728-745 FIGURE 3.8.1-9
WBN 3.8.2 Steel Containment System 3.8.2.1 Description of the Containment and Penetrations 3.8.2.1.1 Description of the Containment The steel containment vessel (SCV) for Watts Bar is a low-leakage, freestanding steel structure consisting of a cylindrical wall, a hemispherical dome, and a bottom liner plate encased in concrete. Figure 3.8.2-1 shows the outline and configuration of the SCV. The structure consists of side walls measuring 114 feet 8-5/8 inches in height from the liner on the base to the spring line of the dome and has an inside diameter of 115 feet. The bottom liner plate is 1/4 inch thick, the cylinder varies from 1-3/8 inch thickness at the bottom to 1-1/2 inch thick at the springline, and the dome varies between 1-3/8 inch thickness and 13/16 inch thickness with 15/16 inch thickness at the apex. The bottom liner plate serves as a leak-tight membrane only (not a pressure vessel). The liner plate is anchored to the concrete by welding it continuously to steel plates embedded in and anchored into the base mat. The anchorage system of the cylindrical walls and the juncture of the cylinder to the base mat are shown in Figure 3.8.2-2. The SCV dome is provided with a circumferential stiffener just above the springline supports, eight penetrations, and several attachments. Two penetrations are for the residual heat removal (RHR) spray system, two penetrations are for the containment spray system, and the remaining four penetrations are spares. The major attachments to the dome consist of lighting fixture supports, header supports for the RHR spray and containment spray systems, and the collector rail supports for the polar crane. Details of these penetrations and attachments are shown in Figure 3.8.2-3. The SCV is provided with both circumferential and vertical stiffeners on the exterior of the shell. These stiffeners are required to satisfy design requirements for expansion and contraction, seismic forces, and pressure transient loads. The circumferential stiffeners were installed on approximately 10-foot centers during erection to ensure stability and alignment of the shell. Vertical stiffeners are spaced at 5° between the two lowest circumferential stiffeners. Other locally stiffened areas are provided at the equipment hatch and two personnel locks. Exterior pipe guides and restraints for the RHR spray and containment spray systems are attached to some of the circumferential stiffeners. During the Unit 1 steam generator replacement, two construction opening were cut into the steel containment vessel. These construction openings were restored by reinstalling the removed steal sections and rewelding them to the remaining structure using full penetration welds. Abandoned-in-place reinforcement and support members added to stiffen the SCV during creation and use of the two construction openings are designed to remain attached to the SCV during a seismic event. The integrity of the restored vessel was verified by NDE and leak testing of the welds. 3.8.2.1.2 Description of Penetrations Most penetration sleeves were preassembled into the SCV shell plates and stress relieved prior to installation of the plates into the SCV shell. Those penetration sleeves which required field installation were provided with insert plates of the same thickness as the shell plates and stress relieved as an assembly. 3.8.2-1
WBN Equipment Hatch The equipment hatch is composed of a cylindrical sleeve in the containment shell and a dished head 20 feet in diameter with mating bolted flanges. The flanged joint has double gasket seals with an annular space for pressurization and testing. The equipment hatch door, sleeve, bolts, and attachments forming the pressure boundary were designed to Section III, Class MC of the ASME Code. The hatch guide system and hatch door hoisting support structure were designed to the AISC Design Specifications. Details of the equipment hatch are shown on Figure 3.8.2-4. Personnel Locks Two personnel locks are provided for each unit. Each lock has double doors with an interlocking system to prevent both doors being opened simultaneously. Remote indication is provided to indicate the position of the far door. Quick-acting type equalizing valves are used to equalize the pressure inside the lock when entering or leaving the containment. Double seals are provided on the doors. The personnel locks are completely prefabricated and assembled welded steel subassemblies designed, fabricated, tested and stamped in accordance with "Section III, Subsection NE" of the ASME Code. Details of the personnel locks are shown on Figure 3.8.2-5. Fuel Transfer Penetration A 20-inch diameter fuel transfer penetration is provided for transfer of fuel between the fuel pool and the containment fuel transfer canal. Expansion bellows were provided to accommodate differential movement between the connecting buildings. Figure 3.8.2-6 shows conceptual details of the fuel transfer penetration. Spare Penetrations Spare penetrations were provided to accommodate future piping and electrical penetrations. The spare penetrations consist of the penetration sleeve and head. Weld caps or closure plates are installed on spare penetrations to maintain containment integrity. Purge Penetrations The purge penetrations have one interior and one exterior quick-acting, tight-sealing isolation valve. A typical purge penetration arrangement is shown on Figure 6.2.3-2. Electrical Penetrations Medium voltage electrical penetrations for reactor coolant pump power use sealed bushings for conductor seals. The assemblies incorporate dual seals along the axis of each conductor. Low voltage power, control and instrumentation cables enter the SCV through penetration assemblies which are designed to provide two leak tight barriers in series with each conductor. 3.8.2-2
WBN All electrical penetrations are designed to maintain containment integrity for Design Basis Accident (DBA) conditions including pressure, temperature and radiation. Double barriers permit testing of each assembly as required to verify that containment integrity is maintained. Qualification tests which may be supplemented by analysis, have been performed and documented on all electrical penetration assembly types to verify that containment integrity will not be violated by the assemblies in the event of a DBA. Existing test data and analysis on electrical penetration types may be used for this verification if the particular environmental conditions of the test were equal to or exceeded those for the Watts Bar Nuclear Plant. Mechanical Penetrations Typical mechanical penetrations are shown on Figures 3.8.2-7 and 3.8.2-8. Mechanical penetration analysis is discussed in Section 3.8.2.4.6. 3.8.2.2 Applicable Codes, Standards and Specifications 3.8.2.2.1 Codes The design of the containment vessel meets the requirements of the American Society of Mechanical Engineers (ASME) Code, Section III, Subsection NE, Winter 1971 Addenda and code cases 1431, 1517, 1529, 1493 and 1768. The design of the bottom liner plates conforms to the requirements of the applicable subsections of the ASME Code, Section VIII, Division 1, and Section III, Paragraph NE-5120. Nonpressure parts, such as supports, bracing, inspection platforms, walkways, and ladders were designed in accordance with the American Institute of Steel Construction (AISC) "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," Seventh Edition. The Eighth Edition is used for shapes not covered by the Seventh Edition. Welding for these nonpressure parts was in accordance with the American Welding Society (AWS), "Structural Welding Code," AWS D1.1 (see Section 3.8.1.2, Item 4). Nuclear Construction Issues Group (NCIG) documents NCIG-01 and NCIG-02 (see Section 3.8.1.2, Item
- 11) may be used after June 26, 1985, to evaluate weldments that were designed and fabricated to the requirements of AISC/AWS.
The anchorage at the containment vessel meets the requirements of the ASME Code, Section III, with a maximum allowable stress for the anchor bolts of 2 x Sm. All containment penetrations including the fuel transfer, purge, and mechanical within the jurisdiction of NE-1140 are designed to Section III, Class MC of the 1971 ASME Code. The penetration assemblies for those penetrations which attach to the nozzles out to and including the valve or valves required to isolate the system and provide a pressure boundary for the containment function are designed to Section III, Class 2 of the ASME Code. Spare penetrations including the nozzle caps are designed to Section III, Class MC of the ASME Code. Two welds (1-074B-D045-01A and 1-074B-D045-08A) in the containment sleeves at the Unit 1 RHR sump have radiographic indications which have been interpreted as exceeding the radiographic acceptance criteria of ASME Section III. TVA has performed calculations (WBN-MTB-025 and CEB-CQS-415) which document the basis for the acceptability of these welds. 3.8.2-3
WBN 3.8.2.2.2 Design Specification Summary Design Criteria The containment vessel, including access openings and penetrations, is designed so that the leakage of radioactive materials from the containment structure under conditions of pressure and temperature resulting from the largest credible energy release following a loss-of-coolant accident (LOCA), including the calculated energy from metal-water or other chemical reactions that could occur as a consequence of failure of any single active component in any emergency cooling system, will not result in undue risk to the health and safety of the public, and is designed to limit below 10 CFR 100 values the leakage of radioactive fission products from the containment under such LOCA conditions. The basic structural elements considered in the design are the vertical cylinder and dome acting as one structure, and the bottom liner plate acting as another. The bottom liner plate is encased in concrete and is designed as a leak tight membrane only. The liner plate is anchored to the concrete by welding it continuously to steel members embedded and anchored in the concrete basemat. On the exterior at approximately 20-foot centers the containment shell is provided with circular inspection platforms which also are designed as permanent circumferential stiffeners. Additional circumferential stiffeners are provided at personnel and equipment hatches and at other large attached masses, along with vertical stiffeners for some distance above and below these attachments. Also, additional permanent circumferential stiffeners were added for stability. Temporary stiffening was not required to meet tolerance requirements specified by TVA in the erection of the vessel. The design provides for movements of the vessel and supports due to expansion and contraction, pressure transient loads, and seismic motion. No allowance is made for corrosion in determining the material thickness of the vessel shell. The following pressure and temperatures were used in the design of the vessel: Overpressure test (1) 16.9 psig Maximum internal pressure (2) (3) (4) 15.0 psig at 250°F Design internal pressure (3) 13.5 psig at 250°F Leakage rate test pressure 15.0 psig Design external pressure 2.0 psig Lowest service metal temperature 30°F Operating ambient temperature 120°F Operating internal temperature 120°F Design temperature 250°F In addition, the evaluations of the vessel design have considered a harsh environment temperature of 327°F. 3.8.2-4
WBN-2
- 1. 1.25 times design internal pressure as required by ASME Code, NE-6322.
- 2. See Paragraph NE-3312(b) of Section III of the ASME Code which states that the "design internal pressure" of the vessel may differ from the "maximum containment pressure" but in no case shall the design internal pressure be less than 90% of the maximum containment internal pressure.
- 3. Typical pressure transient curves are presented in Section 6.2.1. These curves show the transient pressure buildup in the compartments after a LOCA or DBA before a steady-state pressure of 15.0 psig is reached.
- 4. Shell temperature transient curves are presented in Appendix 3.8A. These curves show the shell temperature at the lower compartment wall, upper compartment wall, and ice condenser wall. The maximum containment wall temperature is 220F.
- 5. A postulated main steam line break (MSLB) results in high environmental temperatures (327°F Unit 1 maximum, 325°F Unit 2 maximum) inside the lower compartment of the SCV. However, the coincident internal pressure is lower.[10]
In order to ensure the integrity of the containment, an analysis of the missile and jet forces due to pipe rupture was considered. This problem was eliminated by providing barriers to protect the containment vessel. Typical barriers are the main operating floor (Elevation 756.63) and the crane support wall. An example of a special barrier is the guard pipe enclosing the main steam and feedwater pipes between the Shield Building and the crane wall. Allowable Stress Criteria Allowable stress criteria for the containment vessel are shown in Table 3.8.2-1. The response of the containment vessel to seismic and pressure transient loadings results in a condition in which buckling of the steel shell may occur. Since the ASME Code does not define the allowable buckling stresses for this type of loading condition, an acceptable buckling criteria with appropriate factors of safety is given in Appendix 3.8B. 3.8.2.2.3 NRC Regulatory Guides Applicable NRC Regulatory Guides are shown below. These guides were used as the basis for design of a number of safety oriented features. Regulatory Guide 1.4: Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors A dynamic analysis of the containment vessel was made for the pressure transient loadings. The containment vessel and penetrations were designed to withstand the maximum internal pressure that could occur due to a LOCA and the jet forces associated with the flow from the postulated pipe rupture. Regulatory Guide 1.7: Control of Combustible Gas Concentrations in Containment The containment vessel has a hydrogen mitigation system designed to mitigate the effects of hydrogen releases after a LOCA (Section 6.2.5A). 3.8.2-5
WBN-2 Regulatory Guide 1.28: Quality Assurance Program Requirements (Design and Construction). A Quality Assurance Plan for the Watts Bar Nuclear Plant was developed as a comprehensive plan for the design and construction of the Watts Bar Nuclear Plant. The Quality Assurance Plan of the Westinghouse Electric Corporation, the supplier of the Nuclear Steam Supply System, is also contained therein. The plans were prepared to assure that the control of quality was achieved and documented for each phase of design, material selection, fabrication installation, and/or erection in accordance with the approved specifications and drawings. The plans relate principally to the reactor coolant and safety system, the containment and other components necessary for the safety of the nuclear portion of the plant. The plan assures that:
- 1. Final design requirements and final detailed designs are in accordance with applicable regulatory requirements and design bases.
- 2. Components and systems to which this plan applies are identified and that the final design takes into account the varying degrees of importance of components and systems as evidenced by the possible safety consequences of malfunction or failure.
- 3. Purchased material and components fabricated in vendor shops conform to the final design requirements.
- 4. Components and systems are assembled, constructed, erected, and tested in accordance with the final design requirements and to requirements specified in the UFSAR.
- 5. The as-constructed plant can be operated and maintained in accordance with requirements specified in the UFSAR.
3.8.2.3 Loads and Loading Combinations 3.8.2.3.1 Design Loads The following loads are used in the design of the containment vessel and appurtenances. The loadings for the containment vessel were combined as in Section 3.8.2.3.2. The allowable stress criteria are shown in Table 3.8.2-1. Dead Loads These loads consist of the weight of the SCV, penetration sleeves, equipment and personnel access hatches, and attachments supported by the vessel. The weight of abandoned-in-place reinforcement and support members added to stiffen the SCV during creation and use of the two construction openings made for the Unit 1 steam generator replacement has been included in the evaluated load combinations. 3.8.2-6
WBN Live Loads Penetration loads as applicable. Floor load of 100 psf or 1,000 pounds concentrated moving loads applied to the passage area of the personnel air locks. Construction and snow loads at 50 psf, snow load at 20 psf during construction is considered but not simultaneously with other construction loads. Floor load of 50 psf plus 225 pounds per linear foot for walkways. Thermal Stresses During Design Basis Accident (DBA) The containment vessel is designed to contain all the effluent which would be released by a hypothetical LOCA. This accident assumes a sudden rupture of the reactor coolant system which would result in a release of steam and a steam-air mixture in the vessel. It is calculated that this mixture would cause a lower compartment temperature of 250°F and an upper compartment temperature of 190°F, both occurring essentially instantaneously. After the accident, an internal spray system will commence spraying in the upper compartment only. The spray will discharge water on the interior of the upper compartment. For shell temperature transients refer to Appendix 3.8A. A MSLB produces temperatures in the lower compartment of 327°F (Unit1) or 325°F (Unit 2) with coincident internal pressure and seismic loadings defined in load combinations 3A and 4A. Hydrostatic Loads The containment vessel is designed for three separate flood conditions. Hydrostatic load, Case 1B, accounts for the flooded condition due to ice melt from the ice condenser after the DBA. After all the ice has melted the containment will be flooded to Elevation 719 feet - 3 inches. Also considered is the loading condition during meltdown (hydrostatic load, Case IA). Water will rise to a depth of 2 feet on the floor of the ice condenser. At this time, the depth of water on the containment cylindrical shell will be 9 feet - 3 inches. Hydrostatic load, Case II, accounts for the post-accident fuel recovery condition. In order to remove fuel from the containment after the DBA, the containment vessel is designed for an internal hydrostatic head of 47 feet- 3 inches. For hydrostatic load cases refer to Figure 3.8.2-1. Ice Condenser Duct Panel Loads The outer duct panels of the ice condenser are attached to the containment with threaded studs. These panels impart small horizontal and vertical forces on the containment shell under seismic conditions. The distribution of these loads to the shell is shown in Figure 3.8.2-1. Equipment Loads Equipment loads are those specified on drawings supplied by manufacturers of the equipment. 3.8.2-7
WBN Overpressure Test To test the structural integrity of the vessel an overpressure test of 125% of design pressure is applied under controlled conditions. External Pressure Load The containment vessel is stiffened and designed to withstand an external pressure of 2.0 psig. Seismic Loads Seismic loads are generated using the methodology discussed in Sections 3.7.1 and 3.7.2. Wind Loads The containment vessel and its penetrations are completely enclosed by the Shield Building, and are therefore not subject to the effects of wind and tornadoes. However, during construction, the vessel dome was exposed to the elements for a short duration. For this construction condition, a wind load of 30 psf on the projected area of the vessel dome was considered. Non-Axisymmetric Transient Pressure Loads The division of the containment into compartments is described in Section 6.2.1 and in Section 3.8.2.4.4. Pressure transient loads are considered for occurrence of the DBA (double- ended rupture of the reactor coolant system) in all 6 lower compartment volumes. The curves presented in Section 6.2.1 represent the containment pressure transients for the controlling break locations 1 through 6 for each of the 49 containment elements. The pressures and differential pressures shown on these figures have no margin. The initial containment pressure was assumed to be 0.3 psig. This allows for an initial containment pressure before containment venting is required. The most severe containment pressure differences occur during the first 0.9 second of the blowdown. For structural design purposes the pressures represented by the curves are increased by 45%. This allows for changes in such factors as equipment configuration and openings between compartments, which can influence the flow characteristics of the containment space, the effects of moisture entrainment, and tolerances in the analytical constraints used in the code. (The effects of moisture entrainment, investigated by TVA and Chicago Bridge and Iron Company (CB&I), do not control the design of the containment vessel for any loading condition). Local loadings from commodities attached to the SCV are calculated using dynamic response spectra generated for each area of the vessel. These spectra reflect the response of the vessel to localized dynamic pressure loadings resulting from postulated high energy pipe breaks. See Sections 3.6A and 3.6B for discussions of how these high energy break locations are determined. 3.8.2-8
WBN 3.8.2.3.2 Loading Conditions The following loading conditions are used in the design of the containment vessel:
- 1. Normal Design Condition
- Dead load of containment vessel and appurtenances
- Lateral and vertical load due to one-half SSE
- Personnel access lock floor live load
- Penetration loads
- Design Internal Pressure or Design External Pressure
- Design temperature
- 2. Normal Operation Condition Operating Basis Earthquake (OBE)
- Dead load of containment vessel and appurtenances
- Lateral and vertical load due to OBE
- Penetration loads
- Spray header and lighting fixture live loads
- Walkway live loads
- Personnel access lock floor live load
- Internal temperature range 60°F to 120°F 3A. Upset Condition - DBA and OBE
- Dead load of containment vessel and appurtenances
- Design internal pressure
- Lateral and vertical load due to OBE
- Penetration loads
- Thermal stress loads including shell temperature transients
- Hydrostatic Load Case IA or IB
- Internal temperature range 80°F to 250°F 3.8.2-9
WBN 3B Upset Condition - DBA and OBE
- Dead load of containment vessel and appurtenances
- Pressure transient loads
- Lateral and vertical load due to OBE
- Penetration loads
- Thermal stress loads including shell temperature transients
- Hydrostatic Load Case IA or IB
- Internal temperature range 60°F to 120°F 3C. Upset Condition MSLB
- Dead load of containment vessel and apurtenances
- Internal pressure coincident with MSLB[10]
- Lateral and vertical load due to 1/2 SSE.
- Spray header loads
- Ice condenser duct load
- Thermal load due to temperature range 80°F to 327°F (Unit 1) or 325°F (Unit 2)
- Penetration loads 4A. Emergency Condition - DBA and SSE
- Dead load of containment vessel and appurtenances
- Design internal pressure
- Lateral and vertical load due to SSE
- Penetration loads
- Thermal stress loads including shell temperature transients
- Hydrostatic Load Case IA or IB
- Internal temperature range 80°F to 250°F 3.8.2-10
WBN 4B Upset Condition - DBA and SSE
- Dead Load of Containment vessel and appurtenances
- Pressure transient loads
- Lateral and vertical load due to SSE
- Penetration loads
- Thermal stress loads including shell temperature transients
- Hydrostatic Load Case IA or IB
- Internal temperature range 60°F to 120°F 4C. Emergency Condition MSLB
- Loads are same as in Condition 3C except lateral and vertical load due to SSE
- 5. Construction Condition at Ambient Temperature
- Dead load of containment vessel and appurtenances
- Snow load at 20 psf
- Lateral load due to wind
- Temporary construction live loads on catwalks, platforms, and hemispherical head including support of the first pour of the concrete Shield Building dome.
- 6. Test Condition at Ambient Temperature
- Dead load of containment vessel and appurtenances
- Internal test pressure
- Weight of contained air
- 7. Post-Accident Fuel Recovery Condition with Flooded Vessel
- Dead load of containment vessel and appurtenances
- Hydrostatic Load Case II 3.8.2-11
WBN 3.8.2.4 Design and Analysis Procedures 3.8.2.4.1 Introduction The design, fabrication, and erection of the SCV were contracted to Chicago Bridge and Iron Company (CB&I), Oakbrook, Illinois. The design of the vessel was reported by CB&I in a 12-volume stress report from which the following design and analysis procedures were taken. TVA reviewed the stress report as required by ASME Code Section NA-3260. Furthermore, TVA performed a complete design review of CB&I work to insure the adequacy of the design. As part of the design review, independent analyses were performed for seismic, thermal and pressure transient loading conditions. Compressive stresses in the containment vessel are produced by dead, live, seismic, and pressure transient loads. But pressure transient loads are by far the most significant loads to the stability of the vessel. Therefore, buckling is addressed only in Section 3.8.2.4.4. 3.8.2.4.2 Static Stress Analysis A detailed stress analysis of all major structural components was prepared in sufficient detail to show that each of the stress limitations of the ASME Boiler and Pressure Vessel Code, Section III, Section NE-3000 was satisfied when the vessel is subjected to the loading combinations enumerated in this section. Details of the juncture of the cylinder to the base mat are shown in Figure 3.8.2-2. In the analysis, the juncture was considered to be a point of infinite rigidity. The cylinder at this point cannot expand or rotate under the internal pressure and temperature load conditions; hence, shear and moment are introduced into the cylinder wall. At the point the knuckle is welded to the vessel, a backup stiffener is used. This stiffener gives added rigidity at the point of the weld. Additional protection of the knuckle is accomplished by encasing the knuckle in 'Fiberglass' before floor concrete placement. The embedded knuckle was designed to take interior pressure plus internal or external hydrostatic loads. It was assumed that cracks can occur in the concrete allowing pressure loads on the embedded knuckle. Anchor bolts were post-tensioned to prevent any cracking of the concrete. Thermal and pressure discontinuity stresses in the containment occur one foot above the last weld of the knuckle. The stresses due to dead loads internal, and snow loads were determined at a sufficient number of locations to define the state of stress in the vessel under these loadings. Wind, snow and external support loads on the dome occurred during construction. Stresses due to dead loads, internal and external pressure were determined by hand calculations using classical strength of materials theory. Detail stresses in the embedment region at the base of the vessel were determined from a shell model of the vessel using CB&I computer program 781 described in Appendix 3.8C. The circumferential stiffeners on the embedment region were modeled as horizontal elements and the effect of vertical stiffeners was considered by modeling the shell plate as an orthrotropic material. Forces and bending moments due to the various loads were given by CB&I computer program 781, whereas the resulting detailed stress distribution was calculated using actual geometry of the vessel and stiffening in this region. Design of spherical and cylindrical vessels for internal and external pressure is explicitly treated in Section NE of the ASME Boiler and Pressure Vessel Code. The vessels as designed are in full compliance with the Code requirement for internal and external pressure and provisions applicable to other load conditions. 3.8.2-12
WBN 3.8.2.4.3 Dynamic Seismic Analysis The SCV dynamic analysis is discussed in Section 3.7.2.1. 3.8.2.4.4 Non-Axisymmetric Pressure Loading Analysis The non-axisymmetric pressure loading (NASPL) results from an assumed sudden rupture in the reactor coolant system. The associated pressure loads are dynamic in nature and vary with time in both the circumferential and meridional directions in the vessel. The loads are non-axisymmetric for a short period culminating in uniform internal pressure throughout the containment. For analysis purposes, the containment was subdivided into forty-nine volumes and pressure-time histories determined for each volume for the postulated rupture, i.e., each break in the reactor coolant system. The pressure histories for each of the volumes were computed by the Westinghouse Electric Corporation using the TMD code network documented in Section 6.2.1.3. Figures 3.8.2-10 and 3.8.2-11 show the volumes used to characterize the pressure in the containment. Dynamic analyses were made by CB&I for twelve breaks in the reactor coolant piping, six hot leg and six cold leg breaks. Two separate and distinct analysis methods were used in the design process. The overall vessel response was determined by a dynamic analysis treating the vessel as a lumped mass cantilever beam and by a dynamic shell analysis which considered the effects of local vibration modes.
- 1. Beam Analysis In the CB&I lumped mass beam analysis, each mass represented the mass of the vessel stiffeners and attached masses. The cantilever beam model was loaded with the forces from the NASPL. The forces were resolved into X and Y components and applied as mass point loads in the north-south and east-west directions. The response of the model to non-axisymmetric pressure transients was calculated by CB&I Program 1642 described in Appendix 3.8C. It employs the method of numerical integration and solves for natural frequencies, accelerations, overturning moments, and shears.
- 2. Shell Analyses Independent dynamic shell analyses of the containment were performed by both CB&I and TVA. The shell model used by CB&I is shown in Figure 3.8.2-12. The method of analysis involves a numerical integration technique operating on the governing differential equations. Linear behavior and axisymmetric geometry were assumed. The total transient response was calculated by the sum of the harmonic responses with the input loads being represented by Fourier Series. A full explanation of the method is given in Reference [1]. A number of CB&I proprietary programs, all described in Appendix 3.8C, were employed to arrive at the final shell responses. Figure 3.8.2-13 is a flow diagram of the analysis process with a brief description of the function accomplished by each computer program. CB&I Program 1624 (also in Appendix 3.8C) calculated acceleration response spectra at various elevations and azimuths from the acceleration histories.
3.8.2-13
WBN TVA performed an independent shell analysis of the transient pressure response. A finite element model was used and the solution calculated by numerical integration. The agreement with the CB&I analysis was good. Since the TVA shell analysis was merely a check on the CB&I analysis, full documentation of the process and the programs used are not included herein. The pressures were factored by 1.45 for computing responses to be used to ensure compliance with the buckling criteria in Appendix 3.8B. A factor of 1.80 was used in the design of the anchorage (see Section 3.8.2.4.8). 3.8.2.4.5 Thermal Analysis A thermal analyses was performed on the containment for a loss-of-coolant accident. The shell temperature transients due to a double end rupture of a reactor coolant pipe are described in Appendix 3.8A. The tolerable temperature rise for the steel containment is well above the temperatures shown, since the steel shell was designed for the basic stress limits of Section NB-3221 and Section NB-3222.2 of the ASME Boiler and Pressure Vessel Code, Section III, for ASME SA-516, grade 70 steel at 300° F. Also, as seen by these curves, the containment shell will experience an unbalanced temperature loading for the three compartments. The temperature difference between any two adjacent points on the vessel is held within the limits of Section NB-3222.4 of the code. TVA performed a study to determine the effect of MSLB temperature on the SCV. The impact of the thermal movements on attached penetrations and appurtenances was also accounted for in this study. This study indicated that the SCV and attachments are still within acceptable ASME stress limits under MSLB. 3.8.2.4.6 Penetrations Analysis The vessel manufacturer is responsible for the design of the steel containment including the reinforcement required at the penetrations. The specifications required the manufacturer to submit all preliminary design calculations for TVA's review before any material was detailed or fabricated. Penetrations requiring requalification after CB&I completed their contract were analyzed by TVA. TVA used essentially the same methodology and design criteria as CB&I. However, TVA used its own in-house developed computer program (PNA100 or TPIPE) to sum load combinations and hand calculations to calculate nozzle stresses and a public domain program (WERCO) to calculate shell stresses. The WERCO program employs the methodology of the Welding Research Council Bulletin No. 107. Also, TVA performed an independent analysis of the steel containment, including the reinforcement required at penetrations. Secondary and local stresses at penetrations subjected to applied loads were analyzed by CB&I programs 1027 and 1036, which are described in Appendix 3.8C. These programs employ the methods of the Welding Research Council Bulletin No. 107 in the analysis of the containment shell. 3.8.2-14
WBN Penetrations not subjected to applied loads were designed in accordance with Section NE-3332 of Section III, ASME Code. Most penetrations were preassembled into the containment vessel shell plates and stress relieved prior to installation of the plate into the containment vessel shell. All other penetrations were installed in insert plates of the same thickness at the perimeter as the shell plates and stress relieved as assemblies. As a result, no reinforcement is provided in excess of that available in the shell and neck. Large penetrations, such as the large equipment hatch and personnel access locks, require stiffeners for reinforcement. The penetrations subjected to external loads are supplied with pipe of sufficient wall thickness to resist these loads. Where one or more externally loaded penetrations are in close proximity to another externally loaded penetration or pad plate, the shell was analyzed for the interactive effects of these loaded penetrations. The external loads were assumed to be reversible and the maximum stress combination was determined. Since pressure affects the design of the penetrations, a pressure equal to the internal design pressure is considered to act in conjunction with the externally applied loads. Figure 3.8.2-14 shows the stresses assumed to be present in the analysis of the shell in the vicinity of the penetrations. These assumed stresses, which are due to internal containment pressure, are added to the stresses resulting from the externally applied loads before determining the stress intensities. The assumed stresses are employed as shown in Figure 3.8.2-14 for most of the penetrations. However, it is permissible to reduce these initial stresses when the penetration is provided with greater reinforcement than is required by Section III. At the point of intersection of the shell and penetration, a factor equal to the ratio of the area required for reinforcement within the two-thirds limit to the area available for reinforcement may be used to reduce the assumed initial stresses. At points in the shell away from this intersection, the factor becomes the ratio of required shell thickness to actual shell thickness. This reduction method was used on penetrations which were over-stressed when the assumed initial stresses used were as shown in Figure 3.8.2-14. While the factor for all penetrations using this method was less than 0.5, the minimum factor used in the analysis was 0.5. The neck of the penetration was analyzed using CB&I Program 1392, described in Appendix 3.8C. This program computes the stresses in the neck at two points. The first point is located at a distance from the shell that is outside the normal limits for area replacement. The stresses at this point are due to the external loads and to the containment design pressure acting within the pipe. The second point is located within the area considered for area replacement. In addition to the stresses due to external loads and containment pressure, an assumed stress is also included. This assumed stress is as outlined above at the point of intersection of the shell and penetration and may be modified as discussed above. Permanent caps for spare penetrations are designed in accordance with ASME rules. Flanged penetrations are provided with double gasket details which permit the testing of the gaskets by pressurizing the air space between the gaskets. The Heating, Ventilation and Air Conditioning (HVAC) penetrations were also analyzed by TVA. The entire piping assembly from the flexible connection in the Reactor Building to the flexible connection in the annulus was modeled including pipe, isolation valves, and pipe supports using discrete finite element representation. Shell flexibility was taken into account at the nozzle/shell intersection and at the hanger/shell attachments by inputting equivalent translational and rotational stiffness rates. 3.8.2-15
WBN A response spectrum modal analysis was performed for the seismic and design basis accident condition using the floor response spectra nearest to the penetration locations. The total stress in the nozzle was calculated using the absolute summation of dead load, seismic, and DBA. The stress in the shell was analyzed by inputting the loads above into WERCO. Other non-process and electrical penetrations were also analyzed by TVA. These penetrations were analyzed using the static acceleration technique in which the weight is multiplied by the peak accelerations from the seismic and DBA spectra times a 1.5 amplification factor and applied at the mass center of the assembly. The resulting stresses in the nozzle and shell were calculated using the technique used for qualifying mechanical penetrations. 3.8.2.4.7 Interaction of Containment and Attached Equipment Some items rigidly attached to the containment respond in a non-rigid manner due to the local flexibility of the containment. This effect was analyzed for a number of penetrations and other attachments, but was found to be significant only for the equipment hatch, two personnel locks, and the HVAC penetrations. The following procedure was followed in the equipment hatch and personnel lock analyses:
- 1. Linear and rotational mass moments of inertia were calculated in the radial, circumferential, and longitudinal directions. (The rotational degrees-of-freedom were considered because the centers of mass did not lie in the plane of the containment shell).
- 2. The local stiffnesses of the hatch and locks were calculated for the above degrees-of-freedom. A method developed by Bijlaard[2] was used.
- 3. The periods of vibration were calculated for motions in the radial (push-pull) direction, and in the circumferential and longitudinal (swinging) directions by the equation:
Io T = 2 F K where Io and K are the mass moments of inertia and stiffnesses, respectively.
- 4. The response accelerations for seismic excitation and the pressure transients were taken from the spectra described in Sections 3.8.2.4.3 and 3.8.2.4.4, respectively.
- 5. The total structural response was found by the sum of the effects of the seismic, pressure transient, and dead weight loads.
All of the above calculations were performed by hand. The periods of vibration of the equipment hatch and the personnel locks in the three principal directions were all greater than 0.03 seconds, which is used as the demarcation between rigid and non-rigid vibration. 3.8.2-16
WBN 3.8.2.4.8 Anchorage The containment vessel anchorage system consists of anchor bolts, an embedded anchor plate, and an anchor bolt bearing ring which attaches to the first shell ring. Details of the anchorage are shown in Figure 3.8.2-2. Two rows of 3-1/2 inch anchor bolts are provided with one row on the outside of the shell and one row on the inside of the shell. The bolts in each row are spaced at two degrees and located in pairs on radial lines. The rows are located at equal distances from the center line of the shell. The anchor bolts are embedded in the concrete to the maximum depth available. The majority of the bolts are embedded to a depth such that the lowest point on the bolts is slightly above Elevation 687.0. The remainder of the anchor bolts, located in the area of the pipe sleeves which extend from the penetration for the containment sump, are embedded with their lowest points at Elevation 689.3 being slightly above the sleeves. An embedded anchor plate at the lower end of the bolts is provided to transfer the bolt load to the concrete. The design of the bolt is based on using an allowable stress of 2 x Sm. Allowable stresses in the concrete are based on a specified strength of 5000 psi. Loads considered in the design consist of dead loads, seismic loads, and NASPL loads. The NASPL loads have been increased by 80% for the design of the anchorage. The anchor bolts were pretensioned during construction to assure fixity of the base during an operating accident. Since the concrete is subject to creep over a period of time, the effects of creep were calculated and bolt preload was increased accordingly. The initial bolt strain was calculated based on this preload. The embedded anchor plate is a ring designed to transfer the bolt loads to the concrete. The design assumed that the ring is discontinuous at points midway between bolts. This approach permits the butts in the ring to be unwelded. The tensile loads in the shell are greater than the compressive loads. Since the bolts are preloaded, the effect is that the anchorage is placed in compression. As a result, the anchorage system was designed for the bolt pre-load plus the compressive shell load. 3.8.2.5 Structural Acceptance Criteria 3.8.2.5.1 Margin of Safety A certified stress report was prepared by CB&I for the vessel in accordance with the requirements of the ASME Code. This report contains several hundred pages and therefore is not included in this report. Design values for transient pressure loads were determined by multiplying the calculated values by 1.45 as described in Section 3.8.2.3.1. In addition, the buckling criteria, in Section 5 of Appendix 3.8B, require a load factor of 1.25. 3.8.2-17
WBN Non-pressure parts such as walkways, handrail, ladders, etc., were designed in accordance with AISC "Manual of Steel Construction," seventh edition, so that the stress in the members and welds does not exceed the allowable stress criteria as set forth in the February 1969, AISC "Specifications for Design, Fabrication, and Erection of Structural Steel for Buildings." The factor of safety of these allowable stresses with respect to specified minimum yield points of the material used are as defined in Section 1.5 of "Commentary on the Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings." Local areas, such as the personnel and equipment hatch areas, were checked for deformations to avoid a resonant condition. The vessel as a whole was not designed to deformation limits. Shutdowns and startups do not occur with a frequency to require a design for fatigue failure. The number of load cycles will not affect the containment vessel service life. The stability of the containment vessel was evaluated by the criteria of Appendix 3.8B. This criteria is applicable to stiffened circular and spherical shells and independent panels. A factor of safety was used in the design related to buckling. Loading conditions which included SSE used a factor of safety of 1.1. The factor of safety for external pressure was provided by the ASME Code. The factor of safety for all other loading conditions was 1.25. 3.8.2.6 Materials, Quality Control, and Special Construction Techniques 3.8.2.6.1 Materials - General Materials for the containment vessels, including equipment access hatches, personnel access locks, penetrations, attachments, and appurtenances meet the requirements of the following specifications of the issue in effect on the date of invitation for bids. Impact test requirements were as specified in the ASME Boiler and Pressure Vessel Code, Section III for maximum test metal temperature of 0°F. Charpy V-notch specimens, SA-370, type A, were used for impact testing materials of all product forms in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section III. In order to provide for loss of impact properties during fabrication, all materials were either furnished with an adequate test temperature margin below the minimum NDT temperature, or the specified minimum values were effectively restored by heat treatment in accordance with ASME Code requirements. Material Designations Plate for Vessels Carbon steel SA-516, Grade 70 carbon steel plates for pressure vessels for moderate and lower temperature service. Austenitic stainless steel SA-240, Type 304 Forgings Carbon steel SA-350, Grade LF1 for welding Austenitic stainless steel SA-182, Grade F304 or 316 3.8.2-18
WBN Carbon steel SA-105, SA-181, Grade II, or (for fittings or couplings) SA-234, Grade WPB. Austenitic stainless steel SA-403, WP316, or SA-234, Grade WPB for fittings or couplings) Pipe Carbon steel SA-333, Grade 1 or 6, seamless, or SA-155, Grade KCF70, electric fusion-welded. Austenitic stainless steel SA-312, Grade TP316, seamless, SA-358, Class 1, Grade 316, electric fusion-welded. Carbon steel (for leak chase SA-53 or SA-106 piping and platform handrail piping) Castings Carbon steel SA-216, Grade WCB, or SA-352, Grade-LCB Carbon steel (for lock ASTM A27, Grades 70-36 and hatch mechanisms) Austenitic stainless steel SA-351, Grade CF8M (for personnel lock equalizing valve bonnet, ball, and body) Cold finished steel (for lock ASTM A108, Grades, 1018 to 1050 and hatch mechanisms) inclusive Bar and machine steel (for ASTM A576, special quality, carbon lock and hatch mechanisms) content not less than 0.30 percent. Fasteners Carbon steel SA-320, Grade L7 or L43; SA-193, Grade B7; or SA-194, Grade 2H or 7 Austenitic stainless steel SA-193, Grade B8, or SA-194, Grade 8 3.8.2-19
WBN Carbon steel (for platform A307, Grade B bolts and nuts) Welding Electrodes Carbon steel SFA-5.1, E 70 Classification Submerged Arc SFA-5.17, EL or EM; Gas metal Arc SFA-5.18, E70-S-1 through E70-S-6; Gas Tungsten Arc SFA-5.18, E70-S-1 through E70-S-6. Austenitic stainless steel SFA-5.4 E308 or E309 Classification; SFA-5.9, ER308 or ER309 Classification Structural Steel Plates, bars, and shapes ASTM A36, A283, Grade C, (other than vessel plates) A514 Type F A537, Class 1 Plates (leak chase SA-516, Grade 70. and built-up sections) Plates (platform walkways Regular quality carbon steel and personnel lock floor plate) nonskip S400 Fittings A105, A181, Grade II Gasket materials, including O-ring seals and flexible membrane seals, shall be of Ethylene Propylene Diene Monomer (EPDM) material, Presray Type E603, E603-A or other suitable elastomers in continuous rings and with a Shore A durameter of hardness of 50-70 prior to exposure at operational conditions. Installed seals, packages spares and replacement are to be examined after delivery. Prior to initial startup and then at 18-month intervals thereafter, the installed seals are to be examined. Visual examination is required to determine if there is any evidence of cracking which would result in establishing a leak path for air. If any cracking of the seal is observed, the seal is to be replaced. Minimum values of seal material properties are to be as following: ASTM Before After Spec Exposure Exposure Durometer D2240 50-70 45-75 Min. tensile D412 1800 psi 900 psi Min. elongation D412 400% 150% Max. compression set D095 20% 30% 3.8.2-20
WBN Seals and gasket materials are required to withstand radiation of 108 Rads. 3.8.2.6.2 Corrosion Protection Potential corrosion of the steel containment has been considered at both the embedded bottom liner in conjunction with the concrete, at the inner face in the region of the ice condenser, and at the outer face exposed to the annulus atmosphere. The conditions which determine corrosion are basically the electro-potential of the materials involved, the presence of oxygen and an electrolyte, temperature and may induced electro-potential, from extraneous sources. These have been evaluated in the determination of corrosion. The containment material is to specification SA-516, Grade 70, being a 1% manganese, 0.3% silicon low carbon steel, and has interfaces with concrete. Thus no unfavorable electro-potentials exist in the materials. The climatic conditions for Chattanooga, Tennessee, show an ambient annual temperature of 0°F to 100°F [3]. The corresponding temperature for the steel containment in the region of the ice condenser are approximately 32°F to 120°F. The corrosion of the steel containment face in contact with the containment concrete is not a design consideration since portland cement concrete provides good protection to embedded steel. The protective value of the concrete is ascribed to its alkalinity and relatively high electrical resistivity in atmospheric exposure. Reference [4] identifies three basic conditions as being conducive to the corrosion of steel in concrete.
- 1. The presence of cracks extending from the exposed surface of the concrete to the steel.
- 2. Corrosion cells arising from electro-potential differences in the concrete itself.
- 3. Electrolysis by induced currents in the concrete or steel.
With respect to condition (1) the base consists of a 3-foot thick concrete embedment surrounding all the steel containment. The cracking under the worst of cases is considered minimal. This quantity far surpasses minimum cover recommended by ACI 201-1 in the most corrosive marine environment. The potential for developing corrosion cells was kept to a minimum by limiting the soluble salts and chlorides in the concrete. Further, the continuing corrosion of iron under these conditions requires that the hydrogen deposited at the cathode is freed or combined with oxygen. Since both these mechanisms are prevented by the concrete, the corrosion cells are polarized, and the reaction is brought to a standstill. 3.8.2-21
WBN To preclude the development of induced electric currents and in keeping with good construction practice, all electrical equipment and structures are grounded as determined by the resistivity of the foundation materials for the site. Foundation material resistivity surveys were made and the result considered in the design and determination of the extent of the grounding mat. The seasonal variation of steel containment temperature in the region of the ice condenser gives rise to a range of relative humidity from 4% at 120°F to 45% at 32°F. This is based on saturated air leaking from the cooling ducts at a temperature of 10°F and rising to the steel containment temperature at the containment surface. The annular region exterior to the steel containment is essentially airtight. Only during periods of shutdown during which access doors are open will this seal be broken. In the event of a pipe rupture in the annular region, water would be removed by a drainage system at the base of the annulus. Any ingress of moisture to the interior steel containment face is prevented by sealing the outer periphery of the ice condenser adjacent to the steel containment, and by the vapor barrier on the inside face of the duct panels at the boundary of the ice bed. In the event of any abnormal ingress of moisture through the seal, the leakage air from the cooling ducts has the capacity to absorb moisture up to the limits of the relative humidities quoted above. In addition, any moisture remaining will have a tendency to migrate to the colder end of the temperature gradient; i.e., for all steel containment temperatures above 10°F, moisture will migrate towards the cooling air ducts, where it will be evaporated as the cooling air increases in temperature in the course of its passage through the ducts. For steel containment temperatures below 32°F any moisture at the steel containment face will be frozen, this condition pertaining to relative humidities greater than 45% and steel containment temperatures below 10°F when the migration of moisture could take place from the air cooling ducts to the steel containment. In the event of actuation of the containment spray, water would be applied to the interior surface of the steel containment. Most of the water would be removed by the drainage system and the small amount of moisture remaining would be removed from the steel containment surface by evaporation. Several references have been established which give corrosion data for the limits of the conditions described above. For low alloy steels in any industrial atmosphere long-term tests indicate a maximum total corrosion of 0.016 inch in 40 years (based on 14g/sq dm in 18 years[5]). 3.8.2-22
WBN For dry inland conditions which more closely simulate the steel containment conditions the total corrosion for the plant lifetime is approximately 0.010 inch.[7] This is accounted for by the fact that below relative humidity of 65%, iron oxide itself forms an adherent film affording good protection to further corrosion.[6][8] Furthermore, at temperatures below freezing, ion transport in the electrolyte is almost entirely inhibited, obviating the mechanisms of corrosion.[9] It is concluded that the maximum total corrosion for any exposed internal surface of the steel containment in the region of the ice condenser is 0.010 to 0.015 inch over the lifetime of the plant. In general, the corrosion in the region of the ice condenser is expected to be less than in other areas of the containment, which can be readily inspected. 3.8.2.6.3 Protective Coatings Protective coatings were applied to all exposed steel surfaces of the containment vessel. Surfaces embedded in concrete will not be coated. For coating systems used on the inside of the containment, see Section 6.1.2. As part of the steam generator replacement, two openings in the dome of the containment vessel were created. To support cutting of the openings and reinstallation of the cut section of the containment vessel, the protective coatings on both sides of the containment vessel near the opening cut lines were removed. Areas where the coating on the outside (annulus side) of the containment vessel was removed were recoated following completion of welding, NDE and pressure testing of the containment vessel. Areas where the coating on the inside of the containment vessel was removed were not recoated. These uncoated areas will be periodically inspected as part of the containment in-service inspection program to verify that unacceptable amounts of corrosion of the containment vessel have not occurred. [Historical Information - All exterior vessel shell surfaces and metal surfaces of platforms, floor plate, ladders, walkways, attachments, and accessories located in the annular space surrounding the containment vessel were cleaned in accordance with the requirements of Steel Structures Painting Council Surface Preparation Specification No. 6, Commercial Blast Cleaning, latest edition. After cleaning and having passed inspection, one complete prime shop coat of Carboline Carbozinc 11 paint (dry film thickness was not less than 2-1/2 mils) was applied in accordance with the manufacturer's instructions.] [Historical Information - All interior surfaces of the containment vessel shell and metal surfaces of attachments thereto, except those parts embedded in the base slab and identified as the liner and areas within 2 inches of field-welded joints, were given one prime coat of Carboline Carbozinc 11 within 8 hours after blast cleaning in accordance with Steel Structures Painting Council Surface Preparation Specification No. 10, Near-White Blast Cleaning, latest edition. The primer was top-coated by TVA field forces with an epoxy coating as recommended. The surfaces of the vessel in the annular space were coated with materials selected for the ability to provide protection against atmospheric corrosion.] 3.8.2-23
WBN 3.8.2.6.4 Tolerances The containment vessel as constructed does not exceed the applicable tolerance requirements of the ASME Code for fabrication or erection. The out-of-roundness tolerance does not exceed 0.5% of the nominal inside diameter. The deviation from a vertical line of the vertical cylindrical portion adjacent to the ice condensers is limited to +2 inches for the height of the ice condensers. Threaded studs for attachment of ice condenser outer duct panels do not vary from their theoretical location by more than +1/4 inch. Penetrations do not vary from their theoretical location by more than +1/2 inch. 3.8.2.6.5 Vessel Material Inspection and Test ASTM standard test procedures were employed for the liner and shell plates to ascertain compliance with ASTM specifications. Certified copies of mill test reports of the chemical and physical properties of the steel were submitted to TVA for approval. Tests for qualifying welding procedures and welders were also submitted for approval. All vessel pressure boundary material was tested (one test for each heat of steel) to determine its Nil Ductility Transition Temperature (NDTT). These tests were conducted to meet the requirements of ASME Boiler and Pressure Vessel Code, Section III, Paragraph NB-2300. The tests were conducted at a maximum temperature of 0° F. Ultrasonic inspection was required for all pressure boundary plates subjected to tensile forces normal to the plate surface. This inspection was performed in accordance with ASME Boiler and Pressure Vessel Code, Section III, NB-2530. 3.8.2.6.6 Impact Testing Charpy V-notch impact tests were made of material, weld deposit and the base metal weld heat affected zone employing a test temperature of not more than 30°F below minimum operating temperature. The requirements of the ASME Code, Paragraph NB-2300, were met for all materials under jurisdiction of the code. All weld procedure qualifications for procedures used on the containment vessel shell also meet code requirements for ductility. 3.8.2.6.7 Post-Weld Heat Treatment Field welded joints did not exceed 11/2 inches and therefore, the containment vessel as a completed structure did not require field stress relieving. Insert plates at penetration openings did not exceed 11/2 inches in thickness and stress relieving was not required by ASME Code before or after they were welded to adjacent plates. Post-weld heat treatment, where required, was performed as required by and in accordance with the ASME Code. 3.8.2-24
WBN 3.8.2.6.8 Welding All welding procedures were qualified under provisions of Part A of Section IX of the ASME Code. Welding procedures were submitted to TVA for approval before welding was started. All welding was performed by welders qualified in accordance with Part A of Section IX of the ASME Code. 3.8.2.7 Testing and Inservice Inspection Requirements 3.8.2.7.1 Bottom Liner Plates Test - Historical Information Before concrete was placed over the bottom liner, the leak tightness of this liner was verified. All liner plate welds were vacuum box tested for leak tightness. Upon completion of a successful leak test, the welds were covered with channels, and the channels were leak tested by pressurization to 15 psig. 3.8.2.7.2 Vertical Wall and Dome Tests - Historical Information Welds in the cylinder wall and dome in ASME Code Section III, Categories A and B, were 100% radiographed. Welds in Categories C and D were examined by magnetic particle, liquid penetrant, or by ultrasonic methods. 3.8.2.7.3 Soap Bubble Tests - Historical Information Upon completion of the construction of the containment vessel, a soap bubble test was conducted with the vessel pressurized to 5 psig. Soap solution was applied to all weld seams and gaskets, including both doors of the personnel airlocks. A second soap bubble inspection test was made at 13.5 psig upon completion of the overpressure test in accordance with the requirements of the ASME Code. Any leaks detected by soap bubble test which could affect the integrity of the vessel or which could result in excessive leakage during the leakage rate tests were repaired prior to proceeding with the tests. 3.8.2.7.4 Overpressure Tests - Historical Information After successful completion of the initial soap bubble test, a pneumatic pressure test was made on the containment vessel and each of the personnel airlocks at a pressure of 16.9 psig. Both the inner and the outer doors of the personnel airlocks were tested at this pressure. The test pressure in the containment vessel was maintained for not less than 1 hour. 3.8.2-25
WBN 3.8.2.7.5 Leakage Rate Test - Historical Information Following the successful completion of the soap bubble and overpressure tests a leakage rate test at 15 psig pressure was performed on the containment vessel with the personnel airlock inner doors closed. CB&I performed the leak rate testing by the "Absolute Method," which consists of measuring the temperature, pressure, and humidity of the contained air, and making suitable corrections for changes in temperature and humidity. Equipment and instruments were calibrated and certified before any pressure tests were initiated. Continuous hourly readings were taken until it was satisfactorily shown that the total leakage during a consecutive 24 hour period did not exceed 0.1% of the total contained weight of air at test pressure at ambient temperature in accordance with the requirements of 10 CFR 50, Appendix J. CB&I reviewed the leakage rate data during the test to determine adequacy of the test, authorize termination, or require continuation of the test. 3.8.2.7.6 Operational Testing - Historical Information After completion of the airlocks, including all latching mechanisms, interlocks, etc., each airlock was given an operational test consisting of repeated operation of each door and mechanism to determine whether all parts are operating smoothly without binding or other defects. All defects encountered were corrected and retested. The process of testing, correcting defects, and retesting was continued until no defects were detectable. 3.8.2.7.7 Leak Testing Airlocks - Historical Information The airlocks were pressurized with air to 16.9 psig. All welds and seals were observed for visual signs of distress or noticeable leakage. The airlock pressure was then reduced to 13.5 psig, and a thick soap solution was applied to all welds and seals and observed for bubbles or dry flaking as indications of leaks. Leaks and questionable areas were clearly marked for identification and subsequent repair. During the overpressure testing, the inner door was locked with hold-down devices to prevent upsetting of the seals. The internal pressure of the airlock was reduced to atmospheric pressure and all leaks repaired after which the airlock was again pressurized to 13.5 psig with air and all areas suspected or known to have leaked during the previous test were retested by above soap bubble technique. This procedure was repeated until no leaks were discernible by this means of testing. 3.8.2-26
WBN 3.8.2.7.8 Penetration Tests - Historical Information Type B tests were performed on all penetrations with test bellows and/or pressure taps in accordance with the requirements of 10 CFR 50, Appendix J. See Section 6.2.6 for imposed leak rates and tests performed on penetrations. 3.8.2.7.9 Inservice Inspection Requirements 3.8.2.7.9.1 Components Subject to Examination and/or Test All ASME Code Class MC and metallic liners of Code Class CC components shall be examined and tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and as required by 10 CFR 50.55a(b)(2)(x), 50.55a(g)(4), and 50.55a(g)(6)(ii)(B), except where specific written relief has been requested. The inservice inspection requirements are contained in Section 3.8.5.1.1 for ASME Code Class CC concrete components. The inservice inspection requirements are contained in Section 5.2.8 for ASME Code Class 1 components and Section 6.6 for ASME Code Class 2 and 3 components. Inservice leakage rate tests and inservice surveillance of the containment vessel are discussed in Section 6.2.6. 3.8.2.7.9.2 Accessibility Watts Bar design was established prior to the publication of Subsection IWE of ASME Section XI, however, accessible Class MC and metallic liners of Class CC components will be inservice examined in accordance with the guidelines of Subsection IWE of ASME Section XI. 3.8.2.7.9.3 Examination Techniques and Procedures The examination procedures used by TVA are performed in accordance with the guidelines of Subarticle IWA-2200 of ASME Section XI. 3.8.2.7.9.4 Inspection Intervals An inspection schedule for Class MC components will be developed in accordance with Subarticle IWE-2400 of ASME Section XI and the requirements of 10 CFR 50.55a(g)(6)(ii)(B). 3.8.2.7.9.5 Examination Categories and Requirements The examination categories and requirements for Class MC components will be in accordance with Subsection IWE of ASME Section XI and 10 CFR 50.55a(b)(2)(x) to the extent practicable. 3.8.2.7.9.6 Evaluation of Examination Results Evaluation of examination results shall be in accordance with Article IWE-3000 of ASME Section XI. Components with unacceptable indications will be repaired or replaced in accordance with Article IWA-4000 of ASME Section XI. 3.8.2.7.9.7 System Pressure Tests The program for Class MC and metallic liners of Class CC components system pressure tests shall be in accordance with article IWE-5000 of ASME Section XI. 3.8.2-27
WBN REFERENCES
- 1. Kalnins, A., "Analysis of Shells of Revolution Subjected to Symmetrical and Non-Symmetrical Loads," Trans. of the ASME Journal of Applied Mechanics, September, 1964, pp. 467-476.
- 2. Bijlaard, P. P., "Stresses from Radial Loads and External Moments in Cylindrical Pressure Vessels," Welding Journal 34(12), 1955.
- 3. American Society of Heating, Refrigeration and Air Conditioning Engineers, Handbook of Fundamentals, 1967.
- 4. American Concrete Institute Proceedings, Volume 59 Report No. 201, December 1962, H. R., "Durability of Concrete in Service."
- 5. "Long-Time Atmospheric Corrosion Tests on Low Alloy Steels," H. R. Copson, American Society for Testing Materials Proceedings, Volume 60, 1960, pp. 650-666.
- 6. "Corrosion," Metals Handbook, Volume 13, ninth edition, Metals Park, Ohio, 1987, pp 82-83.
- 7. Lauobe, C. P., "Corrosion of Steel in Marine Atmospheres," Trans. Electro Chemical Society, Volume 87, 1945, pp. 161-182.
- 8. Rozenfield, I. L., "Atmospheric Corrosion of Metals," Houston, TX: NACE, 1973, pp 104-106.
- 9. Evans, Ulrick R., The Corrosion and Oxidation of Metals: Scientific Principles and Practical Applications. London, Edward Arnold (Publishers) Ltd., 1960, pp. 27-37.
- 10. TVA drawings 47E235-44 through 48, "Containment Harsh Environment."
3.8.2-28
WBN TABLES 3.8.2-1 ALLOWABLE STRESS CRITERIA - CONTAINMENT VESSEL Material: SA-516, Grade 70 Applicable ASME Code Reference Loading Conditions for Stress Intensity
- 1. Normal Design Condition NB-3221
- 2. Normal Operation Condition NB-3222
- 3. Upset Operation Condition NB-3223
- 4. Emergency Operation Condition NB-3224
- 5. Construction Condition NB-3221
- 6. Test Condition NB-3226
- 7. Post-Accident Fuel Recovery Condition NB-3224 1
All references are to the ASME Boiler and Pressure Vessel Code, 1971 Edition, Section III.
1-8' tCPN(;) L J*J ~ a a'
- av 0
<<1 Ii wall too' II 7TP/CA[ CORNER
$ea/r J'*Y-0, 3'thiek l0yer 7W ArZW ftmVpx)betameen leakage ebonne/s and on rodvrr of Arakage cbsnne/s,*ohberr.
WT4 ACTU: ALL MATERIAL SHALL BE FABRICATED AND ERECTED IN ACCORDANCE WITH THE AISC CODE. This d/eamy is a dFkoe of 5rpueyah Neclear plant drawirlp 4BN401 RS except as noted by encirclements on RO issue. Elevations not encircled on RO issue are Segvoydh Reactor Owd np a/eretroms raised byplas ZJ=0. Encircled e/rvdfions on RO issue have of been changed otberusa to satisfy design requirements. W.P. denotes work point lie contractor SAO// divide Hose h'net plate irate four indrprndent test area.. A// a/ovahen. and dimensions air computed to the mood. of the containment she//, Yn/e.s noted butts s~.v ZC r., sections and detai/o not shown see 4910402 4 P/am see 49840/ Material by TVA denotes material to be furnished and
/ ea &Wt. fabrieated by TVA fir/d.
All rt,vctu,.l *fret furnished by TVA field 1. bedSTMA-.?f. All pipe furnished by TVA field to be AS TN A/e0 or 4", except as noted.
~En,eow /ydiose+, All headed concrete anchors to be .4STM 4/08 E/actrodea All holds made Ay 7YA on ,temr other than Pre"wre boundary parts to be n aecordanee w th geoera/ construction specification 0-Z9.
Reae tot IuJdinp is a class 1 structure Welding d00emfntat/en And inspection by TVA field to be in actordence with construchm tpec(hcatir. 1035-11/. A// /esAape thoane/s no ..pier to be encased with Owens Corm.np 'Fiberp/ar'or epvat. UN type V3 or TYA RF/O9C fberg49 os eW cwwd cr Nvegv/r aA*ma ovd ame " 705 ar 704 PF100 tberyas os err tXW swfiees. Procurement,fabnc0,06 and decomentotien ofmaferial by contractor shall confirm to the TVA OA plan. QA. for all material by TYA field to be poolity level 1 per WBRP eanstrvct/on .pert Ficalren N3i-19/ EAN. QA. tbral/rjdtrmrl dy TAAfTehf hsMgva/ty Awl/2rpor W,6#P eonJhaetiao spe cificoh-w Alm-8W&,.w0/any/. FOR HATCHED AREAS SEE UNIT 1 AC DRAWING 48N401 WATTS BAR FINAL SAFETY ANALYSIS REPORT REACTOR BUILDING UNIT 2 STRUCTURAL STEEL CONTAINMENT VESSEL ANCHOR BOLT PLAN &. BASE DETS SH-1 TVA DWG NO. 2-48N401 R1 for bokApe L. a Co 400 secliomr 8 0.4ils FIGURE 3.8.2-2(U2)
Awe: Z Iwo("( I
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- R DETA/L 6 (°(
R/.S. 3Q - /iyhf/oy fixture and a max r~ DETA/L D efM=O:ste 809 00/. i / 7YACA4 LATNr/Na AUrURE 9UPPafr Lee" ono, qow Scah /("-/=O' f Collector foil
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DETAIL A /,ten/lit seyport padr,rnisfirt EcbsyrbeoiseNeilArArAG/7dI. Z. Cexdvit sr,pert pods H be gvahte lore/x inaccord"nce with APKP v , COLLECTOR RA/L SUPPORT SECT/OIV A-A nnatrvcfien sporho'canon AJS-ee/.
- Ge&h'&o d Nt P/M nyptrN .p 1 Sca/t 4'r-1,0' Sca/e /1'=/*
1 0" 3.Onisf-Vf,ciaaePNK/and Mir I febt94101ity/tut/D'/n i ~~ see Ot ail O 0CCerdat4Ce with IYOKP ten&trvrfitn pterficafien NJO*Sa/. 41,fir rvrfise voting tf everh000,and ver//ea/eend41it supports of /' //iyhtiny tohduif vet va/etrot c/omp
- e. Amw areoo or qer/ with a max epociu, . 1 N=0'.
DETA/L C mor AYnY ~Nw*n) s.This drawing A a d4 w*cofo o/ sogvoytb kFFii 0.1114;
/ °0x7°Ig B ryp quadrant (9OhbH0) 40~*R2 areopPos /4~,& &whv/smeaf! M R0 issue.
I R,g '! sails hcatren '. petail 4. 6.E,kvatWA notencircled on MAW* We Sop wyea RAwAr evi/diny DErA& A/ headed oono J awr(typ) rail wppor/s,rea Oefti/ ckwtiwea /v/'sod ly dvs E3=0" Encircled tlewtion3 ore RV issue A.Tw bore Dean changed otharvase fo satisfy dhyn regw ameats. 7.for gevarrt note& SM 4VANO/ S.Pipe by o*hw, , w/pbo AJA Schsside Ab, AM 11104 WM b full of wafer: Akzz[t wieht por pipe w1N be 6:5*/0`0k tsowac DOME PL AAA IMMI OR ArrAC/WiVrB &"W M w, L/CMTIW P/yr Ac RA/L 3AOM J e l!o/*e*0 We swab r*2o* Note 0:Ste akago Ordte &no s~ E/ 66.,_6* i2;10'.11p e !f&n 41,00ovs(mrdAwt E[ myvIez98%Sw MCa/-7S320)for spray headtr and cWarfor CJ rail details. Qf7A/L O e' I 7N'-e'9r 2lE7CT'621030~e MK /0 0 B-B As ahoen 6rwrlM
/£ub rypl7l /pNf Scab N~C'rKa 1
1 7-940' I TVA Pisld lomoke I I e/fechtwnf Ab I/yhtkV cl pixfnfo end cable cbffvm
/7J7' At t.9:l7' Maximum we4hf eg1w M 'Wopar a41pparr y
I e 20/,d rylA Oo/hc!/sy pips by erhtrs. Ksse/-' E 8"oRY44Q elptey Aersire r Pbr boa#se ssr Dome i C;oe a Saet E*EA~F. Cbnlalnmtnt ,frray A7* psr e.30/:f17<JOb' 28'9'R'to t conAW RNR Jpry O pQeets ~.t76*SJ06;bydNefo C-C.' Cj
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scok r're ~ ~~P%Po(eyofaerl~
e *b DETA/L D Mrre dA-wflowW pips rratminf. - - rYP/GL Nt PIPE sUPMRr pipe p/Ms Aar/zonfa/ rwstre/nt. ? SCALE:I"-/10'
/E' A06 4-Aoorfr.
AOrE E: FOR HATCHED AREAS SEE F/ELD MAY GSE TrAVAL DEWL Ax YJt* UNIT 1 AC DRAWING 48N404 Az t9J*O'oboro FOR CMWlr 8W%0Mr8 /ASIQE - *: r: :*.R [kt E/EI0.!t0" P WONNAL MR LOCKS. Sys Az tse*/J'be/ow Mork:RATES mo mtx4rbE WATTS BAR satyr/rurm mR pt. /NOETML ;S r Qu ffAz2J0J0'exvoRt Sysf.rAt4z 24rJ7'0EL 7Jlz0* l rt FINAL SAFETY (
`E/e32c/t' ANALYSIS REPORT CROSS SEC7/C7W K x io x is MN tale aWFEAPOe A"29000W= awnrtt oeRY C dsTE aSO REACTOR BUILDING N4W UNIT 2 E 2°Rad(tyn) STRUCTURAL STEEL Ab#b A dEr=f A&t/r see Now 000 ofP- Pbta A rrra,.el DETA/L C E-E CONTAINMENT VESSEL
~aes6/
Untetrut d CI~q K Untstrut R Clamp QE7711L F INTERIOR ELEVATION SHEET 2 K-/f >gale-AWs G-6 TVA DWG NO. 2-48N404 RO Egahrf"+/w' FIGURE 3.8.2-3(U2)
11 Ladder
.~w~~~X Abre yoisf Axat/on to be EL. 6'06=ti Two speed eA*ciric hoist
!_ 6'reac,or Okrg ~1 adiustab/e to suit ono hu'sr support by N door c.g. covsfoinment resse/
G'Onlractor. yoisf cvnlro/s DETA/L A C-C M be mounted on inside or Lbnlainment she// wa// Sce/e /i'./--O" and accessible from f/oor V. 756'-7i" Ie0° ,- _ -f/asl xcess f/eNorm KEY PLAN Safery Aook _ Sce/e /"=30=0` L-~D EL. 7P7=6" D -D yo/JT ACCESS PLATFORM Inside of hatch NOT SyOWN
£`O-rhg seals h; Retch in raised and ili dogged positive. // ~\
~i Ori// and faP for test fitting. Test Ca ction to be Mcwted 0/0rt accessible ,noinf outside of/AVlch.
DETAIL B ll ~~ Sca/e 6" -/-O" Do dewKe for991
-Z-914-1 'age I. ONE HATCH PER UNIT REWIRED; TWO TOTAL
- 2. WATCH TO HAVE 20'-0" DIAMETER CLEAR OPENING.
!. HATCHES DESIGNED FOR LOADS, PRESSURES AND TEMPERATURES, thers LOAD COMBINATIONS, AND STRESSES AS OUTLINED IN SPECIFICATION OR SNOWN ON THIS DRAWING.
. ALL LIFTING DEVICES SIZED FOR S,1 MIN SAFETY FACTOR,
- 5. DESIGN FOR PRESSURE OF 13.3 PSIG AND TEST PRESSURE OF IGI SIDE OLEAKTITAIENT L.
- 6. HATCHPSHALLNRE"IN GNT WITNEA NEGATIVE PRESSURE OF 0.5 PSIG INSIDE CONTAINMENT VESSEL.
- 7. ALL MATERIALS AND WORK BY TVA FIELD SHALL BE IN ACCORDANCE WITH GENERAL CONSTRUCTION SPECIFICATION NSG-BB1, QUALITY LEVEL II.
/ug ing guide Sce% 1"= /'-O" Excepf as noted WATTS BAR Auxihbry FINAL SAFETY FOR HATCHED AREAS SEE ANALYSIS REPORT UNIT 1 AC DWG 44N250 REACTOR BUILDING UNIT 2 EQUIPMENT ACCESS HATCH ARRANGEMENT & DETAILS SECT/ON 6-8 TVA DWG NO. 2-44N250 RO ELEVATION A-A FIGURE 3.8.2-4
NOTES: L ITEMS /L1ENT/F/ED XX ARE TVA MAR,AM. 47MT00-A2'X CLASS AS 50"MA/ED IN THE ¢70A4600 S6R/LS.
- 2. m DENOTES THE RECOMMENDED SEQUENCE FOR FIELD MELDS.
I S. NY: DENOTES NATERTIGHT. S4'-R- TO A. SS: DENOTES STAINLESS STEEL. ff REACTOR S. FUEL TRANSFER TUBE AND ATTACHMENT MELDS ARE CLASS E,
~WT ASNE CODE, SECTION III. TVA CLASS B.
I/W731.2 m m" 2"EIR4NS/AN R JOINT
.1 r*t k' \
\
/) 8T U2 ONLY
- 6. TO BE FABRICATED IN ACCORDANCE KITH TVA CONSTRUCTION SPECIFICATION NO. G-29.
F. FUEL TRANSFER TUBE ALIGNMENT TO BE AS SPECIFIED BY 4 EXPANSION JOINT / NESTINGHOUSE FIELD ENGINEER.
** m e. All MrIER/AL BY FIELD rO BE 4240 TP 301 S.s. EXCEPT k AS NOTED.
t! WT t1,~1 ~*'~ *./ ,\ VCSEALS 4/#7/1.3 '`*i ~* *.`. **' Um t ONLY ' .*' e* 9. A// MATERIAL BY FIELD TO BE OA LEVEL 1. EXCEPT AS NOTED. FLEXIBLE SEAL *' * - 47B473-92 t** + 10."SOME WELDS SHOWN ON THIS DRAWING (SERIES) WERE FOUND TO BE rm *~ *E t. EXPANSION BELL >S ASSY MK NO 47W455*/0 DEFICIENT BY THE WELD EVALUATION PROJECT,EVALUATED BY WATTS BAR A ENGINEERING PROJECT AND FOUND TO BE SUITABLE FOR SERVICE.SEE DESIGN Vii.' ~** CALCULATIONS FOR EVALUATIONS'. REFER TO RIMS NO.820890901319 k I'4" 1 21'4 00. 110'4"
*.*.~ *t. .~
onn n AND ALSO PIRWBNMEB8889. M BY FIELD. Sf'VINT 1---a f F7/fL TRANSFER 77NE EC 7//2~1~ WS83F690 _ _ 15 77 EL 709.1! - 2o*aD. - INNLL 7IAC, DETAIL D 1r-~ J PENIM TO ANNUL US BELLOW~MANSIB N A Ue ATAF O EL 7ME'-B" 48#406 MK X-3 ~*'
- 2'-iD I CARBON STEEL PIPE ya-DRILL IN BOTTOM O*
MST CONVOLUTION. ' 1 SJI"STA/NLfS9 STEEL SCN 405(BY F/ELD TMNMENT LIAtR ~,D '1 ** *~' **. *t SEE 48#935.6 * . O wrm O7 ONLY n 2B:ss*z9'45 Wr rV'8'TO 6 REACTOR FOR BOLTING MATERIAL, SEE FOR 3LEEYE . 8'-1"TO NK p OR ORIGINAL BILL MATERI LOCATION SEE WTU7 ONLY 47BU488-1 OR SA 584 TYPE e 94 471,470-3' S CONTAINA/ENT OR
-~BN~os PLAN VIEW-FUEL TRANSFER TUBE B LOCATE SUFYORT AT PENETRATbN /
L(N/T / AS SHOWN UNIT C OPPOSITE NAND / **221S'O D *10k ID BY naD U7 ONLY TK4 F/ELD 710~7r AAD SUPFO.fr LL T/3.l1' N7! /N ACCIMMACE MAN 47AOSS SER/F9 T7/RONCN ACCESS LLOCr AREA(Armcmo 777 SHIELD WALL ONLY) f EL 718'-1" CLASS CUSS B' 1/2'3000* SS. HALF _ WT COUPLING, SA/82F186, (SEe OW6 47AT) BYFIELD. DRILL 5/6* TNROUGN. /. JCL r Fm 7AWFER V4S!'BDV1S
,'WL TAAAGYER b 7WE IA SECTION D*D pAl*Y ~~N~*/- N- T. S.
SECTION A A 6.6'PLATE ROILED BY FIELD TO NAAE 21"0A NAE TNREADSO CAP BY FTELD SHOWN MU SEAL REMOVED SECTION C-C NTS HIS 4'ED
- 22Fi'00.,BY FIELD 20.
SS BAR%OD, ROD AM 07, By FIELD VSo NOS BYAWS SECTION B-B HIS EL Li"Ld0 21.957"va/D 0 d FbEL TRANSFER TUBE FUEL TRANSFER TORE SECTION E-E NMI WATTS BAR 7--
}~PATHRAY LLOHS (DRA9INC MN780-1)
FINAL SAFETY A6"PLATE ROLL£D 1'0 MARE 24'AD , NAEJ BY FIELD 24. O.D. REF. } REFERENCE DRAWING=: ANALYSIS REPORT WT m We$T/NG NOOSE: 26* C.D. REF.
'1' .583F690 - FUEL TRANSFER TUBE DCTA/LS TE WTm 95'2161 -"SrINGN09$9 E Spec SST LINE "r p' PLATE 1/4"SST BY FIELD SST PT PLATE 1/0" SST BY FIELD POWERHOUSE
- W3 ' S REF TbA 47EM4SS-/- I'M 7WfNSFER nag 4/LL OFMATEMAL AUX REACTOR BLDGSUNITS 1 & 2 Z) 78
- 78 DETAIL DETAIL F
+s,M447Lq.2i47455iLDAM TRANSFGROTUBE &Z~ MECHANICAL OFrl00WrML L&W NTS UNIT 1 ONLY NTS UNIT 2 ONLY 47BN600*SEABf!-Mtt7MMfNT2Gb/VTXOL B/LL OF MA72MAL FUEL TRANSFER TUBE SEE DETAIL F FOR UNIT 2 INSTALLATION FOR BELLOWS ASSY SEE TVA CONTR WK64-83233, PAMWAY RMS I-D722*,2-D722
- TVA DWG NO. 47W455-1 RL FIGURE 3.8.2-6
PENETRATION RITHOUT BELLOWS RATION DESCRIPTION /NOCESS PIPE VESSEL FENETRATION FWEO NFID LOCATED ON TYPE T.Y.A. 00 CORTAffAME TVESSEL 31*+2. 1 - K' O-D, 2 - 2 Iaa//ows o/nf *tr4rn6lt/0R O.D.A. 'CIA W1.OX CI YACN W.7RIC ASME SI2E ME N0. ap00/ piece 04- pftm 'F~ W/A2 " 6 -333 e Type Mr- pfs.n C m% YI/s/d ant. YII X - 16 RORAM CHARGING 330' 2.882' ISO 150' 0 .66* BART 6t i S AR 47W406 rype W.oatt we At ISIrnrAor seeA// .eeA1 w 7R -3 6 X X-15 INST LINES a1A8' - 1 660' 60 6X8gE Defy/ F2* T"w $-oaad A4- socket wa661IF ' X 61 Y.978' 40 6A 6 481;_a's .Gts- 6.094* BO 6X 3/ 6! 8.60 47woleXY/ CNILLEO WATER -_ f AV w/stcA process P406 AOL FrAw zw- Aw" wen/ ants -De/w/FI YI X-41 FLOOR p6MfP SUMP D/SC// 2.578' - IK 200- 1 Fled we YII X- 42 I" MAKE - OF 3.50*J 403 300* 300
'A6ad(A51 YI X-434 SEA. WATER INJECTION 2.378 - 6 / O Slow 6Xt 470 If0 sp*/pe4e Cgs)
X- 43G -
-~j~ c<<AAra*,ebf ~a X-43C IqK/ v X-4110 DFTA/L A7 X - 77 DEN1411794L12ED WATER
-31 180* 100 47849 TYPE Jzr f Ev AW A4- samir OfIA/L DI 8 REVEL PURIF PUMP SYCT 6.et5* GASO 403 Type S Ar mew IstAe as prow" Novas AA,. Ire// matAr66 Goes, ! 10.
Type Im for doff wom So paces AMMO VII X-49 REVEL W ROIRIF ftWDISC 4.50* 4AM 403 for 080 8.685* 0.010* 0X 4 47V454 Hof M enAt 160 X X-64 IAG RVLIS 4-1A6* - I50E6 6.iEi* 6094* 6X1/e! q S7L6 , ro X-36 INST LINES 4-1.Or - 180 8Ad7! 850' 2406 8 -3 4711600 6.618* 6.094* REACTOR A xv/ x- 63 COILLEO MGTER 2.373" - 40 46- LS 6.623.6.014* 60 Ox3/6R 47WlMi 6'6 b 6 owPEI m xvl x- ee - 1 * - 3b'RG OVwt
.h.~ /=6(TTP~~ NM/ Ame1 Yll X"44 RC T 4.50' 4. 403 150- 200 8.613 0.020 00 SA 6 X4 4711406 Dated E I Deter/ A2 YI x -6 CONTROL AIR aim - 4011 1 150- 020 4.Sr 4.044 00 4X t 47600
! b Allow 700 3 80. _ 6 47W6OO X X - N INST LIMES 1.06* - 160 900' 3 X3/6E -
Dale!/ F2 I-cwm Rif ti x - 94 RADIATION MONITORING 3-1.N 406 6.625* 6.094- 00 6X3/8 At 478800-103
~' wAvN4V1 X - !! RADIATION MONITORING 47x600-1OS 6idY~ > 1-366X-26 INST LINES 3-.318. - lea /50- 30 4711500 eesR Aesd CYlN7A/NMENT~ * \. AIM.
X -r INST LINES 1*o - + 650* 2465 AMOCO PENE7RA7ADN Aroceu ppe ContoRRleRt 8Aa12 ISO Llo SAM 8.020* .3 NOZZLE YI X - 34 CONTRDI. AIR .A6* - 4DS S B X2 47N600 ArRet/vNon 1 ~ J YI11 X-46A AUXILIARY FEEDWATER 4.50
- 3.80 00 600* ties 6AY6* 6.094* S -J S 6 X 1 - 4711427 TYPE xr VIII X-408 AUXILIARY FEEDWATER 4.60* 3.916! 80 "30 600- 6A26- s- axe 7 TYPE Ml TYPE= 10 IX X-4AA COMNN6l21T 8PRAY 10,73* 10.01'0 40A TVP304 1la' 180 I/.0" 19.809* 100 i4 X 10 GRF304 4711437 Far butt weld C:S. peers Naas AOT 717 SCALE X-40 CONTNA AE NT aFRA' 1o.7r 10.070* 409 14.0- I9AOT 100 14X 10 Net AD awe X -41A RNR BpRie 40S / nor 1276* 12A63* 12 X S 150 X-488 RNR~ 8.625 0 On- fYA69* 12X S YI X - !7 INST LINES 1,05* - 160 nor 2235 S.S. 3.061" X - 3X3/4 2-47!100-89 woN Tom= SSG!I00 At 00 X X - Be INST LINES - 9X3/4 2-47/800-89 001`811/ F2%~ DeWA6'~ ~iD~F2 F Ym.w.C79X!"/!f> Ali X -!! Hp ANAL SAMPLE 1.31 - 3 I 47!825-11 SA'31 8II XY/ X 67 CMILLCD eMTER t.e 40 49* Ia b.b 6.014
- 60 6A 6X3/6R 0840 47we16 Jo- 1901 REACTOR RYIcoeW.elero IB* 17.30. .31` T f90
' 10 14-* 3.303* 0,374- l 24x18 47W492 X
XTA/L a ;< DETA/L Of X11 X-/s 11* 17.30.'1 .ass* TP304 19T AR 1r* 3.303* 0.876' 24x16 4711492 Sao". Se6k1:J* -A4-a 612 TO ACC. 1.315° - 180 160- app Mg'eCN! p -- let/ Atel(3.a.) I Q , XIII X - as M1 To PN 1.05 - 408 312 ISO- I50 10.750 /QO AO 471VWO 7F0.8r60p !p-*/ /rest(as) ~10 X 3/8 TX 4-ep CaNRwrNRSRf SPARES (2) /.OS - 408 iP8D4 - NOTE: SPARE 3/4° PENETRATION PIPING IS b P* etrRha/ meek CAPPED ON GOTH ENDS (X-390 AND X-39D). for 8x9/Aix 47W435-22
-t4- Al 1-110 SPARE 2.375* - too 2736 6.027 5.820-A 1/8' DIAMETER HOLE IS DRILLED IN THE END OF THE CAP ON THE INSIDE OF THE GOV. IYI X-114 GLYCOL FLOOR COOLING 2.375* - 40 -f0' 190 20- It 20x3/8 Tx 4FW482 FD19 wMLC 7MOMIF3'i(7O~YKLI/S2*E -
3/4° WELDED XW X-115 GLYCOL FLOOR COOLING 40 '/0' ISO 20 /9-war e0 x 3/8 Tx 4711482 1-IL TYPE zr
/ac as n0ee6as 6Ave11 r PIPE CAP (TYP) 3/4° SPARE vl X-11 LONrwL AIR 1.370* -
2.375* 408 150- 120 4.SO' 1.04 d0 4A 1 4711800 (SEE NOTE) N*t h Yak X X-92A SPARE 1.050° - 180 _ - - 3.3' N/A 180 -333 3°X 3/4' - - 7NK A&+0tow 3/4° WELDED X X-929 SPARE 1.030° - iGO _ - - 3.5' N/A 180 - 3°X 1/4' - - leseh sop" A4M l AjA LC e*f PIPE CAP (TYP) N FABRICATION MOTE: (SEE NOTE) X X-92C SPARE 1.050° - 110 - - 3.8' N/A is0 - 3°X 1/4'Ow - 04 - nrvrrn* mr**we X1/32 UND. skp+ ~ R AM 1`A -*-~S TAI N/A 3°X 3/4' Ows-PAM¢ tG AM X X-92D SPARE 1.030° - iB0 _ - - 3.S' 180 - X X-100 INST LINE 1.050° - 180 853- 2235 3.S' 3.081 XX-STG 47!825-11 rw 1wN/ f6rt6wtas. 0to " sSLEEVE NOTES: DFTA/L E2 Are, Reyes sad "ference dxy's see 47W33/-/. Aw! AP ac" MR MMU MQUESS 6RbRTER7MMN,4* /NA AID TYPE 3w I119LL JX crmaf AAelt/ line penstret/en x-!! Id of DFTA/L G2 jC1 Not to agkL CawtRAl..aAt CDNL. AIYT70SC4LE /rear lMldl( / 0 -) 3/4' 88000 OR /3000 SW CAP 3.988° ASME SA182F TP 304 BY FIELD 3° DIA SJ6tve /SEE 47W432) (TYP) MK 47TOW-226 (NP) ?I: CQNTA/NA/ENT 3'-0' - 1'-0' INSIDE CONTAINMENT FOR X-25C, 3/4° 88000 S.W. CAP * *p l *- BUMP 9-1/2°(REF X-BSA, X-BBC, X-86C, X-86D a ASEBi SAl WY TP 30+ r3/4° X-92D BY FIELD INSIDE I OUTSIDE CONTAINMENT FOR X-23, B CONTAINMENT FOR X-238, X-88A, X-888, X-57A, 120. X-28 /2-16 - X-92A, X-828 a X-92C OUTSIDE CONTAINMENT 'e /.. NOTE: FOR PENETRATION X-92A a A D FOR X-106 a X-106 X-928 USE MK NO. 47W625-A888 O UNIT 2 ONLY - SPARE - OR EQUAL (3000 OCAP) pe 6 UNIT 2 ONLY -SPARE ~W 8R1 WATTS BAR
- 1. 4-3/4° SCH 160 PIPES SPACED 80 APART FOR *APPLIES TO X-86 a X-92 FINAL SAFETY 1-1/2'0 CM PIPES NOT GOR4 ro CLARITY PENETRATIONS X-25.X-27.
X-84,X-85,X-88,X-87 TYPE g MULTIPLE LINE PENETRATIONS
.** 6
- A
** *9
*
- F?
(FIl90W!/De) ANALYSIS REPORT (~) T SCH 180 PIPES SPACED (LOOKING FROM OUTSIDE THE SCV)
- 8°0 SCH 80 PIPE 12P APART FOR "F^ P/RDW90 SA333 PIPE'~DETAIL GR.B PENETRATIONS X-26 A X-96 ACHINE~ I IM6MIA.X18 (EXIST)
DETAIL H2 3. 3-1-1/2° SCH 40S PIPES SPACED 12P APART FOR ~: MARKS 478Mf433-A8271 1-1' 180 PIPE CENTERED C 1-3/4° SCH 180 PIPE CENTERED S N POWERHOUSE FOR X-99 PEN X-94 AZ 294, EL. 741' PEN X-98 AZ 293, EL. 741' PENETRATIONS X-94 a X-95
- 4. 1-3/4° SCH 180 PIPE A 1-1 '83000 S.W. CAP ASME SA 182, pi P304 UNIT 2 ONLY PL O.D.-2.858' FOR X-100 UNIT 2 ONLY ELEN07/DAl
': REACTOR BUILDING NOTE. ENGINEERING APPROVAL SHALL 13E REQUIRED PRIOR TO ATTACHING CENTERED FOR X-93 (C O.D.
NOTES: NTS
- 1. STRUCTURAL PLATE SHALL CONFORM TO SAS18, GR 70.
*x(TR ADDITIONAL INSTRUMENT TUBING LINES TO THIS PENETRATION. e 2.988°) a UNIT 2 X-98.
INSIDE CONTAINMENT UNIT 2 ONLY PL O.D- 2.856° ~~ UNIT 2 S. 1-1° SCH 160 PIPE SPARE
- 2. STRUCTURAL 1-1/2°e SCH. 40S PIPE SHALL CONFORM TO SA312, TYPE DETAIL K2 PENETRATION WM-2-PENT-0304-0026 ONLY. CENTERED FOR 2-X-99 (E MOT10801RE MECHANICAL 304SS. FULL 2.938') e TYPE % CONTAINMENT PENETRATION
- 3. EXISTING 3/8° THE. PLATE CAP ASSEMBLY 91ALL REMOVED AND THE SCV PENETRATION PREPARED AS SHOWN IN EDGE °A° PRIOR TO 7WF Y NOTE: FIELD TO SUPPLY SCALE: AS NIOT£D Instrvw.ew Peff fartib w n INSTRUMENT PENETRATION INSTALLATION.
- 4. ALL MATERIAL AND FABRICATION BY FIELD. Alf h saa/k MATERIAL FOR PEN.
X-86 a X-87, COYReMOAI OR:ePMIG& 4)IV391-1 TVA DWG NO. 2-47W331-2 R1 2-X-98 A 2-X-99.
- 3. ALL MATERIAL SHALL BE FABRICATED AND ERECTED IN ACCORDANCE WITH ASME SECTION III. 14, S/1 FIGURE 3.8.2-7(U2)
0 z 3 Q fr C3 W. AWEDESIGN L. THK ASH@ C1 *OUCH* W. THK AGE ASME p THK MAT'L TEdP. F PRESS OR SOH DESIGN PSI OR SMHL MAT'L I.D. OR SCH MAT'L MAT'L MIN. SA-185-1701 1.012' SA-788 51.303° SA-518 SA-390 47/400 p MIN. / KCP 70 MIN. L KC 70-1 OR 70 QRLF2 W 47/400 Z 47/400 Q h- 47/400 Z SA-333 29.303° 47/401 N QR8 Q 47/401 47/401 p Q 47/401 U SA-376 27.303° SA-182 47/432 TP-316 678'304
.438° ATP-500441 YBA03' SA-1 47/432 W30482 SA_ as 15.303* SA-333 SA-390 47/400 OR S OR 8 QRLF2 SA-312 17.303' W-333 SA-192 47/408 TP 304 OR a W304 SA378 19.303' SA-333 SA-182 47/438 TP 304 OR 8 OW304 SA_ as SA-333 SA-380 OR 8 18.303' OR a ORLF2 47WSSD SA-378 23.303° SA-333 SA-792 47/438 8.628° TP 318 OR 8 018'304 8.628° g TP-378
" 318 23.303- 47/438 SA-370 SA-518 TP 318 19.30* OR 70 47/438 SA-378 SA_333 TP 318 19.303* OR 8 47/438 SA-012 19.303' 47/488 TP 304 3.80' 12 17.303* SA-333 IF 304 OR 8 47W880 O 633 19.503' ORLF20 47/482 SA-333 47/482 OR$ 19.303' ~F2 0 SA-312 12.063' 12 X e % 3/4 ~~ 47WSSD TP 304 SA-390 1 - 1 30 X 24 X 14 OWy~ 47/432 RHR RECIRC SUCTION TYP 304 SA-398-t 55 R HR RECIRC SUCTION TYP 304 =0sh-0-1 30 X 24 X 14 Wye 47/432 CONT. SPRAY RECIRC TYPJ04 1 29 X 20 X 12 WS0 4 47/437 CONT. SPRAY RECIRC S A~tW 1 20 X 2D X 12 ~81 O04 47/437 SIS PUMP DISCH TO HOT LEG SA-378 19.303* SA-1 TP 316 G -6J 20 X 14 X 4 47/433 SIS KW DISCH TO COLD LEG S TP`-376 378 19.303' A-~ 20 X 14 X 4 W ' 04 47/433 SA-476 23A03* 24 X 18 X 12 Wy04 NONE 47/435-22 TP 318 SA-370 23.303° 24 X 18 X 12 ~y82 NONE 47/435-22 TP 318 1S.SEE DRAWING 4SN406 FOR PENETRATION NOTES:
LOCATIONS 1. FOR SPECIFICATIONS SEE SPECIFICATION FOR PENETRATION 17.PER 168281 DOCUMENTS HISTORICAL ASSEMBLIES, TVA SPECIFICATION 2529 CONTR. 76K81 83180 NONCONFORMING CONDITION REPORTS (NCR.) AND 2. FOR GENERAL SPECIFICATION SEE TVA SPECIFICATION 1521 SUBSEQUENT INSPECTION REPORTS ON THE ALLOW CONTR. 74C38-83018. TYPE PENETRATIONS (SCRATCHES, DENTS, AND/OR 3. CONTAINMENT VESSEL PENETRATION NOZZLES BY OTHERS. GOUGESNX-7, NSA. 2X-88, 2X-8D, 2X-12A, 4. PENETRATION ASSEMBLIES SHOWN FOR UNIT 1, UNIT 2 ARE 2X-12B, 2X-12C, 2X-12D, 2X-13A, 2X-138, IDENTICAL. 2X-13C. 2X-13D. 2X-14A. 2X-148. 2X-14C. 5, PENETRATION WITH BELLOWS REQUIRE THE PROCESS PIPE TO HE 2X-14D. 2X-17, 2X-21. 2X-22, 2X-24. 2X-3O. INSULATED BY FABRICATOR DURING: ASSEMBLY WITH MATERIAL 2X-32, 2X-33, 2X-48, 2X-48, 2X-47A, 2X-476, AS PER SPECIFICATION. THE INSULATION SHALL INCLUDE ONLY 2X-81, 2X-107. THESE DAMAGES AREREFERENCE. ZINC; THAT PORTION OF THE PROCESS PIPE WITH IN WARD PIPE AND DOCUMENTED AS A NOTE FOR FUTU~ SHALL EXTEND 1° BEYOND THE END OF THE GUARD PIPING. 18.PER 482739-001 DOCUMENTS A NONCONFORMING S. BELLOWS OUTLINE IS SHOWN FOR ILLUSTRATIVE PURPOSES. CONDITION WITH SLEEVE R2C208. THIS SLEEVE CONTRACTOR TO DETERMINE THE LENGTH AND NUMBER OF HAS BEEN EVALUATED AS ACCEPTABLE AS IS AND CORRGATION. U S NO FURTHER ACTION IS NECESSARY. THIS 7, WIDE RINGS ON OD OF GUARD PIPE FOR PENETRATIOBNS X12A, B, CONDITION IS BEIN: DOCUMENTED AS A NOTE FOR C. D. X13A, 8, C. AND D AND DEFLECTION RINGS ON ALL FUTURE REFERENCE. PENETRATIONS TO BE FIELD WELDED IN PLACE AFTER GUARD PIPE 19.11HE FOLLOWING PENETRATION MARK NUMBERS HAVE IN INSERTED THROUGH VESSEL PENETRATION NOZZLE. HISTORICAL. BELLOW MISALIGNMENTS: 2X-2OA; 8. TYPES III, IV, AND V ANCHORED IN THE SHIELD BUILDING WALL. 2X-2069 2X-21; 2X-22; 2X-24; 2X-30; 2X-32; 9. PENETRATION ASSEMBLIES WILL HAVE A PREFIX INDICATING UNIT 2X-33; 2X-13A; 2X-138; 2X-130; 2X-13D; DESIGNATION, I.E., UNIT 1, 1-X-16, UNIT 2, 2-X-18. R 2X-14A; 2X-148; 2X-14C; 2X-126; 2X-12C; 10. CSH DENOTES, 'CONTRACT STARTS HERE.' SPLIT ANCHOR 2X-12D; 2X-SD. ACCEPTABILITY QF THESE TT. BELLOWS FOR K-14, K-15, K-18, AM K-17 HAVE THE FOLLOWING S/8°(MIN}- SLEEVE NONCONFORMANCES WAS DOCUMENTED WITH PERM X-108 ONLY 143565, 143533 (SUPPLEMENTEDWITH PER DESIGN CONDITIONS: REP. OR 1058158 413817) AND 143588. MAXIMUM INTERNAL PRESSURE - 27 PSIC MAXIMUM EXTERNAL PRESSURE - 25 PSIG 2O.CONSTRUC SECTIONS OF GUARD PIPE ORIGINALLY 12. ALL WELDS TO BE SHOP WELDS, UNLESS OTHERWISE NOTED. CONSTRUCTED FROM SA-188 KC 70-1 CAN REPLACED WITH ROLLED PLATE CONSTRUCTED FROM 13.FIR C WELD NUMBER [~} ETC. DESIGNATE FIELD WELDS. SEE 48N408 21.rTAC SA 315 GR. 70 OR SA 518 GR. FOR CONTAINMENT TO PENETRATION FIELD FI LEEDS. SEE 1.WARD PIPE IS AN INTEGRAL ATTACHMENT TO THE APPLICABLE DRAYO DL63 FOR PROCESS PIPE FIELD LEEDS. PRESSURE BOUNDARY, BUT NOT PART OF THE 14.FLUED HEAD EXTENSION SPOOLS MUST ~ COUNTERBORED A MINIMUM PRESSURE BOUNDARY. DISTANCE OF 2T WHERE IDENTIFIED AS REQUIRING INSERVICE INSPECTION. 15.DEFELCTION RINGS MAY BE SPLIT IN HALF TO FACILITATE INSTALLATION. THE 2 HALVES ARE NOT REQUIRED TO BE WELDED TOGETHER. WATTS BAR FINAL SAFETY ANALYSIS REPORT POWERHOUSE REACTOR BUILDING-UNIT 2 MECHANICAL CONTAINMENT PENETRATIONS TVA DWG NO. 2-47W331-1 R1 FIGURE 3.8.2-8(U2)
FIGURE 3.8.2-9 DELETED
N E W S UFSAR Amendment 1 WATTS BAR NUCLEAR PLANT WATTSFINAL BAR SAFETY NUCLEAR PLANT FINAL ANALYSIS SAFETY REPORT ANALYSIS REPORT TMD Nodal TMD NODALVolumes VOLUMES FIGURE 3.8.2-10 FIGURE 3.8.2-10
WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT WATTS BAR NUCLEAR PLANT FINAL SAFETY TMD NODAL VOLUMES REPORT ANALYSIS Figure 3.8.2-11 TMD Nodal Volumes FIGURE 3.8.2-11
Z0.10.625
~- °94.62s
~ 1949.25 ,n 166 ; ai 14 u
/3 .62542. o i 238. 6zS Q3 9(o.0 rr~4.czs 114.0 /d/o.6 ZS r.0.0 6a 0- PRES. CHANGE LockTlaws ro~o. ~Zs 3t 54.0 Q- NUMBER OF SE4ME.NT5 9s~.6zs Y
- a. f- DIM6NSIows TO THF. RIQ&4,T 8 2. 62S 2 .p ARE LE W4THS OF EACH PkwLl-(U 81E.62S O 90.0 4-728.4 72.0 DIMENSIONS TO THE. LEFT AIRE DISTANCES ABOVE BASE(.N s7Sbzs CIRCUMFERENTIAL STIFFENEkS a 3 7S.o 10"x I~
7625 1 3o.0 I FE 13/F* 457.625 = I- 22"x 89.0
¢ ~ Y a VERTICAL STIFFENS R-26o.625 6.50 37* C ix 61/
164.6Z5 127.625 SPA C IAlq( 7Z 4RO UIVZ) I15.6ZS WATTS BAR NUCLEAR PLANT I2. FINAL SAFETY WATTS BAR NUCLEAR PLANT ANALYSIS REPORT FINAL SAFETY ANALYSIS REPORT CBI CONTAINMENT SHELL MODEL Figure 3.8.2-12 CB&I Containment Shell Model FIGURE 3.8.2-12
Model Areas of Shell Having Vertical Stiffening on Program E781. Save Resulting Shell Stiffness Matrices. Using Program E1622, Calculate Fourier Amplitudes of the Input Using Program E1372, Combine Pressure Versus Time. Stiffness Matrices from Above with Those Generated for Shell Sections With No Vertical Stringers & Save. Also Calculate Natural Frequencies. Using Program E1374 Determine the Response in Each Harmonic Using Program E1623, Sum the Harmonics, Calculate Maximums, & Store Required Data for Buckling Check. WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT CB&I Containment Shell Analysis Flow Model FIGURE 3.8.2-13
...__ .- Assn M ~~ sT a.~s5 CtQd,uM MER.iv a po I ry T i e-YL 5PN CYO SPH s r~
A suRFA.e c iyz 5 I'/zsml?/zsM t'~2s 3s MsMe9?,os.f.1E 5m SM M 8 MEMBRANE S,M Sw, /LSm Sti,1.15 SVQf= ACE Sw, Srn S Srn S I C mEMBVA ME $rn 15 m Y2Sm 6h 5vol.4 4wPC-NisTQA-noN WATTS BAR NUCLEAR PLANT T FINAL SAFETY ANALYSIS REPORT STRESS REDUCTION METHOD WATTS BAR NUCLEAR PLANT FINAL SAFETY FigureANALYSIS REPORT 3.8.2-14 Stress Reduction Method FIGURE 3.8.2-14
WBN 3.8.3 Concrete Interior Structure The concrete interior structures are designed as described in Sections 3.8.3.1 through 3.8.3.8. The evaluation and the modification of the interior reinforced concrete structures are optionally done using the ultimate strength design method in accordance with the codes, load definitions, and load combinations specified in Appendix 3.8E. 3.8.3.1 Description of the Interior Structure 3.8.3.1.1 General This structure, shown in Figures 3.8.3-1 through 3.8.3-7, is a complex assemblage of reinforced concrete walls, slabs, and columns housed inside the steel containment vessel (SCV). It will act as a temporary containment while routing steam to and through the ice condenser in the event of a loss of coolant accident (LOCA). The reactor, four steam generators, four reactor coolant pumps, pressurizer, ice condenser, reactor instrumentation, air-handling equipment, and various other support systems are located inside this structure. The portion of this structure which separates the upper compartment from the lower is defined as the divider barrier (See Figure 3.8.1-1). The failure of any part of the divider barrier is considered critical since it would allow LOCA steam to bypass the ice condenser, thereby increasing the pressure within the steel containment. For this reason the divider barrier is designed more conservatively than the rest of the internal structure. Since the ice condenser is both a structure and an engineered safeguard system, most detail information can be found in Section 6.5. 3.8.3.1.2 Containment Floor Structural Fill Slab The containment floor slab is a reinforced concrete slab of 3-foot nominal thickness cast on top of the bottom liner plate. Reinforcement is provided in both faces to withstand uplift pore pressure below the liner plate and to develop restraint for uplift and rotational moments at the base of the crane wall. Earthquake shearing forces are transmitted to the base slab through shear keys located below the crane wall, through a direct tie with the reactor cavity, and through direct bearing on the base of the Shield Building wall as a result of the expanded volume of the fill slab under operating temperatures. Stresses resulting from shear forces are very low since any one of the three methods is capable of transmitting the entire shearing force. The interior concrete structure is sufficiently keyed to the reactor cavity by its configuration of walls and slabs to provide base stability against earthquake overturning moments above the bottom liner plate at Elevation 699.28. In addition, the anchorages for the steam generators and reactor coolant pumps supports tie the containment structural fill slab to the base slab in the vicinity of the crane wall providing additional stability. 3.8.3-1
WBN 3.8.3.1.3 Reactor Cavity Wall This 17-foot inside diameter. circular wall supports and encloses the reactor vessel above the lower reactor cavity. The wall is 8-1/2 feet thick, primarily for radiation shielding and structural requirements due to the reactor support loads, and it extends from the base slab at Elevation 702.78 to Elevation 714.96 where it intersects the refueling canal floor slab. Neutron detector windows reduce the effective structural thickness to 6 feet for approximately the first 10 feet of height. The next 12 feet of height has only a 4-foot, 3-inch structural thickness due to the 3-foot, 1-inch wide by 6-foot, 6-inch high inspection cavity which surrounds the reactor vessel. Between the inspection cavity and reactor vessel is a 14-inch-wide structural wall. This is shown in Figures 3.8.3-7a through 3.8.3-7g. 14-Inch Reactor Cavity Bulkhead Wall This wall consists of two interconnected concentric cylindrical steel shells, separated by 11 inches of concrete fill, that form the reactor cavity from Elevation 721.625 through a composite steel diaphragm to the concrete wall at radius 13.0. This is shown on Figures 3.8.3-7e through 3.8.3-7g. The anchorage for the wall at Elevation 715.04 is the reactor support embedments directly below the wall. Both the 14-inch wall and the diaphragm are designed to withstand the pressure and temperature transients resulting from a LOCA condition in accordance with the required factored loading combination in Table 3.8.3-1. The peak differential design pressure and temperature are 50 psi and 150°F, respectively. 3.8.3.1.4 Compartment Above Reactor This compartment is approximately a 270° arc continuation of the reactor cavity wall. The ends of the arc intersect the two refueling canal side walls. The inside diameter of the wall is 26 feet and the thickness is 4 feet. It extends from approximately Elevation 725.15 to the bottom of the divider deck slab at Elevation 754.13. This compartment is vented to the lower compartment area, outside the wall, by six windows which reduce the wall to five columns. These columns each have a cross sectional area of 12 square feet and extend the last 4-1/2 feet of height to the bottom of the divider deck slab. This compartment is shown in Figure 3.8.3-7b. During reactor operation this compartment is sealed across the top by the concrete and steel missile shield and is sealed across the refueling canal by a concrete and steel gate. Seals Between Upper and Lower Compartments The seals extend across the gap between the inside surface of each steel containment vessel and the concrete structure within each vessel. They are located along the bottom of the concrete floor under the ice condenser, at Elevations 739.5 and 751.33 between the ends of the ice condenser and the refueling canal concrete structure, and along the vertical sides of the refueling canal structure. These seals form part of the barrier between the upper and lower compartments of the containment vessels. 3.8.3-2
WBN The seals consist of long strips of flexible elastomer coated fabric folded longitudinally with open edges butted and sewn to form two loops in cross section. Metal bars are inserted into the seal for use during attachment. These strips are field-spliced with vulcanized overlay joints or cold bond overlay to form a continuous seal. The seals are attached to the containment vessel and the interior concrete structure using bolted clamps with bolts spaced one foot apart. These clamps grip the metal bars inserted in the seal thereby closing and sealing the gap. These seals form part of the barrier between the upper and lower compartments of the containment vessels. During normal operating conditions, the seals prevent airflow around the ice condensers. In an accident, the seals and the other divider parts limit the amount of hot gases, steam, and vapor that can bypass the ice condensers. The seals will maintain their integrity for the first 12 hours after an accident. A small amount of leaking during this period is permissible. The seals will maintain their integrity during earthquake conditions and effectively maintain their air seal. The seals will function effectively in a post-earthquake condition. The slack in the coated fabric seals, which was purposely provided, allows for the relative movement, between the containment vessel and the interior concrete structure, which results from earthquakes. 3.8.3.1.5 Refueling Canal Walls and Floor (Divider Barrier) These irregular shaped walls and slabs vary in thickness and enclose an area approximately 19 feet by 32 feet. This area will be filled with water along with the compartment above the reactor during refueling operations. The water level will be about 35 feet above the canal floor slab. The reactor internals will be removed and stored in the refueling canal during refueling. Refueling canal walls and floor are shown in Figure 3.8.3-6. 3.8.3.1.6 Crane Wall This basically 3-foot-thick, 117-foot-high cylindrical wall encloses an 83-foot inside diameter area containing the reactor, reactor coolant pumps, steam generators, and reactor coolant piping. The crane wall is 4 feet thick in some areas to satisfy structural requirements. There are four localized areas of "bumps" on the wall which are 4 feet thick. These "bumps" begin at the floor slab at Elevation 702.78 and extend to the Elevation 716.0 and are approximately 25 feet in arch length. The crane wall is also 4 feet thick between Elevation 737.42 and Elevation 756.63 over its entire circumference. This wall acts as the major support for the divider barrier slabs and walls. It also supports the floors and walls in the 13-foot annulus between it and the steel containment vessel (SCV). The 175-ton polar crane is mounted on top of this wall. Over the refueling canal the wall has a section removed leaving a curved beam 23 feet deep spanning an arc length of 41 feet between ice condenser compartment end walls. Beginning at Elevation 746.42 the crane wall has twenty-four 7-foot, 4-inch high by 6-foot, 8-inch openings for the ice condenser inlet doors. The remaining wall consists of 25 columns each having a 10-square-foot cross section. Above the operating deck floor at Elevation 756.63, the crane wall is part of the divider barrier. It is also part of the pressurizer and steam generator compartments, which constitute part of the divider barrier, and is designed to resist the same pressures as these compartments. At the top of the crane wall the steel support beams for the ice condenser bridge crane cantilever over the ice beds causing moments and forces in the crane wall. Lateral seismic loads from the ice beds are transmitted to the outer face of the crane wall. 3.8.3-3
WBN Personnel Access Doors in Crane Wall See Figures 3.8.3-8 through 3.8.3-11. Four access doors in the lower half of the crane wall are provided in each Reactor Building at the following locations: Floor Elevation Azimuth 702.78 221° 702.78 299° 716.0 114° 16 11 745.0 299° The doors provide passageways 3-feet wide by 6-feet, 6-inch high through the concrete crane wall for workmen and tools. When closed, the doors seal the passageways against steam jets, pressure, and missiles that may originate from pipe rupture in the compartment inside the crane wall. Each door is manually operated and hinged to a steel frame embedded in the concrete wall. Each door consists of a steel skin plate stiffened by horizontal framing. The skin plate is faced with a cushioning structure of vertically arranged square, steel tubing separated from the doors skin plate by a collapsible latticework of steel bars, the purpose of which is to absorb the energy of missiles striking the door. The cushioning structure is covered with sheet steel for appearance. Bearing of the door against the frame is through steel bars. An elastomer seal is attached to the periphery of the door to reduce the possibility of damage from jets to items beyond the door. Two lever-type latches operable from either side hold the door in the closed position. Hinges on the doors are provided with graphite impregnated bushings. The doors, under normal operating conditions, provide an effective seal against airflow and can be operated and secured manually from either side. For pipe rupture accidents, the doors seal the passageways in the crane wall against missiles, jets and pressure that may originate within the crane wall enclosure, thus preventing consequent damage to the containment vessel and to piping and machinery between the crane wall and containment vessel. The doors will maintain their integrity and seal for not less than the first 12 hours following an accident. Limited leakage during this period is permissible. All parts of the doors, except the seals, are fireproof. Increased leakage may occur during a fire. It is assumed that a fire and an accident which require sealing will not occur simultaneously since the reactors will be shut down immediately if a fire develops. 3.8.3-4
WBN 3.8.3.1.7 Steam Generator Compartments (Divider Barrier) Two double-compartment structures house the four steam generators in pairs on opposite sides of the building. Each structure consists of curved and straight sections of walls that vary in thickness from 21/2 to 4 feet. Divider barrier walls around two steam generators extend 42 feet up from the divider floor and are capped with a 3-foot-thick slab spanning over the steam generators from the crane wall. A wall between the two steam generators extends from the divider barrier walls to the crane wall, completing the double compartment. The center wall extends only 321/2 feet above the floor. The area above the top of this wall, except for that occupied by a beam acting as a barrier for a postulated break in a pipe, will reduce the compartment pressure buildup in a single compartment by venting the steam to the other compartment. See Figures 3.8.1-1 and 3.8.3-6. During the Unit 1 steam generator replacement, four construction openings in the steam generator compartment concrete roofs were created to facilitate removal of the old steam generators and installation of the replacement steam generators. The compartment roofs were restored by connecting the removed section of concrete to the remaining structure using bolted splice plate connections. The gap between the removed concrete sections and the remaining structure was filled with non-shrink grout to restore the integrity of the divided barrier. 3.8.3.1.8 Pressurizer Compartment (Divider Barrier) This compartment separates the pressurizer from the upper compartment. Its walls project about 38 feet above the Elevation 756.63 floor where they are capped with a 3-foot-thick slab. It is similar to the steam generator compartments except its wall thickness varies from 2 to 3 feet and the volume is much smaller. See Figure 3.8.3-6. 3.8.3.1.9 Divider Deck at Elevation 756.63 (Divider Barrier) This 21/2-foot-thick irregular shaped floor is the major divider barrier between upper and lower compartments. It is supported at its outer edges by the crane wall and the compartment walls for the steam generators and pressurizer. Support near the center of the building consists of the refueling canal walls and the five columns of the upper reactor compartment. This floor contains five hatches for equipment removal. The concrete covers on these hatches are designed for the same loadings as the floor. The floor outline is shown in Figure 3.8.3-3. 3.8.3.1.10 Ice Condenser Support Floor - Elevation 744.5 (Divider Barrier) This floor extends 12-feet, 8-inches from the outside of the crane wall to the 4-inch expansion joint separating it from the steel containment vessel. A circumferential beam under its outer edge is cast with the floor. This edge beam is supported by concrete columns which extend down through the Elevation 716 floor to the fill slab at Elevation 702.78. The floor extends 300° around the outside of the crane wall between the ice condenser end walls at azimuths 245° and 305°, as shown in Figure 3.8.3-1. 3.8.3.1.11 Penetrations Through the Divider Barrier Canal Gate The canal gate consists of three removable concrete wall elements as illustrated on Figures 3.8.3-2 and 3.8.3-4. The elements are 2-feet, 6-inches thick and span between 7-inch-deep slots formed in the walls of the refueling canal. 3.8.3-5
WBN Control Rod Drive (CRD) Missile Shield The CRD missile shield consists of three removable concrete slabs as illustrated on Figures 3.8.3-3 and 3.8.3-4. The slabs are 3-feet, 6-inches thick and are anchored to the divider barrier slab at Elevation 756.63 by anchor bolt assemblies. The details of the anchor bolt assemblies are shown on Figure 3.8.3-14. Reactor Coolant Pump Access and Lower Compartment Access Access to the reactor coolant pumps and lower compartment is provided by removable slabs as illustrated on Figure 3.8.3-6. The reactor coolant pump access slabs are approximately 10 feet in diameter and the lower compartment access slab is approximately 6 by 10 feet. Both are 2-feet, 6-inches thick and are anchored to the divider barrier slab by anchor bolt assemblies around the edges. The details of the anchor bolt assemblies are shown on Figure 3.8.3-17. Equipment Access Hatch This hatch consists of a removable structural steel framed hatch cover and a structural steel support frame adjacent to the containment vessel. The arrangement and details are illustrated on Figures 3.8.3-6 and 3.8.3-18. The support frame and hatch cover consist of structural steel wide flange sections covered with steel plate. To provide adequate seals between the upper and lower compartment, the side of the frame adjacent to the containment vessel was designed to span from the refueling canal wall to the divider barrier slab, a distance of approximately 5.0 feet. The hatch cover is anchored to the concrete structure by anchor bolt assemblies at each end of the cover. Escape Hatch The location of the hatch and the details are shown on Figure 3.8.3-12. The hatch consists of a frame embedded in the divider barrier floor with a hinged and manually operated cover consisting of skin plate stiffened by framing. Quick-acting wheels are provided for opening and closing the cover from either side. Coil springs are incorporated with the hinges to reduce the force required for opening the cover. The hatch is equipped with a limit switch which operates to give an indication in the control room of the position of the hatch cover. Air Return Duct Penetration The air return ducts penetrate the divider barrier at two different locations as indicated on Figures 3.8.3-2 and 3.8.3-3. One penetration is at Elevation 746.0 and the other is at Elevation 756.63. The penetrations are 4-foot, 6-inch (inside diameter.) circular openings with flanges on both sides to provide attachment for the ventilating ducts. The details of the penetrations are shown on Figures 3.8.3-15 and 3.8.3-16. 3.8.3-6
WBN Pressurizer Enclosure Manway The location and details of the manway are shown of Figure 3.8.3-20. The manway consists of a 30-inch diameter sleeve embedded in concrete at Elevation 798.0 at the top of the pressurizer compartment. The manway cover is a circular steel plate that is bolted in place to provide adequate sealing between the upper and lower compartment. 3.8.3.2 Applicable Codes, Standards and Specifications Structural design of the interior concrete structures is in compliance with the ACI 318-71 Building Code Requirements for Reinforced Concrete, and ACI-ASME (ACI 359) Article CC 3000 document, "Standard Code for Concrete Reactor Vessels and Containments." Reinforcing steel conforms to the requirements of ASTM Designation A615, Grade 60. [Historical Information - Installation, inspection and testing requirements for plain and reinforced concrete used in the construction of Category I structures, as well as for fly ash used as an admixture in concrete, were in general accordance with the ASTM standards, ACI 318-71, ANSI N45.2.5 and Regulatory Guides 1.15 and 1.55, except for the following TVA specific requirements:]
- 1. Required Qualification Tests - Historical Information
- a. Fly ash - TVA uses its own specification for fly ash rather than ASTM C 618.
Significant differences occur in requirements for fineness, pozzolanic activity index, and loss or ignition. ASTM C 618 has two requirements for fineness. The first, a surface area obtained by an air permeability apparatus, is not conformed to by TVA. The second, the amount retained on a No. 325 sieve, is conformed to by TVA. TVA's requirement for pozzolanic activity index with portland cement is 65% of the ASTM C 618 requirement, but TVA's procedures result in substituting fly ash for fine aggregate in a mix and thus increase the quantity of fly ash available for reaction with the cement. TVA's limit on loss on ignition is 50% of that in ASTM C 618. The most recent addition to ASTM C 618 is a limit on the product of loss on ignition and the amount retained on the No. 325 sieve. This was added when a statistical analysis indicated that it correlated with the effect of fly ash in concrete. TVA's limits on the individual items will result in conformance to the ASTM C 618 limit on the product. TVA's experience at hydro, fossil, and nuclear plants indicates that their specification for fly ash produced acceptable concrete.
- b. Water and ice - TVA complies with the suggested limits of CRD C 400 rather than the suggested limits of AASHO T-26. The suggested limits on compressive strength of mortar are the same. AASHO T-26 utilizes an autoclave soundness test developed specifically to test free lime or magnesia in cement. ASTM C 150 has a specified limit on autoclave expansion of 0.8%, but many elements exhibit less than 0.1%. The ASTM analysis for precision indicates that repeat tests by the same operator can differ 21% by their means. This illustrates that a clear indication of unsoundness due to water will be difficult to obtain. The test for soundness is not recommended in ASTM STP 169-A "Significance of Tests and Properties of Concrete-Making Materials."
3.8.3-7
WBN
- c. Concrete mixes - TVA does not conform with two of the recommendations of ACI 211. The recommended limiting water-cement ratios were developed for Type 1 cements. ACI 211 recommendations do not agree with ACI 318. Where fly ash is utilized, neither can be directly applicable. The recommendation that trial batches for strength be made at maximum slump and air contents should not be applied where statistical analysis establishes an average over-strength requirement. Use of maximum air content and slump will offset the average strength and invalidate the analysis.
- 2. Required In-process Tests - Historical Information
- a. The construction procedure for this project is in substantial agreement with ANSI N45.2.5 frequencies for those tests required by TVA.
- b. Mixer-uniformity - TVA's requirements for unit weight of air-free mortar and for coarse aggregate content are more restrictive than ASTM C 94.
- c. Compressive strength - The sampling frequency for compressive strength provided by ANSI N45.2.5 appears to be intended for a transit mix operation.
TVA purchase specifications for ready-mix are even more restrictive, however, the vast majority of TVA concrete is produced in a central mix plant where the provided frequency appears excessive. TVA varied the testing frequency requirements based on the specified strength of concrete with no one sample to represent more than:
- 300 cubic yards for a specified strength of 2000 psi.
- 175 cubic yards for specified 28 day strengths of 3000 psi or more.
These test frequencies were in effect until January 1978 during the majority of concrete placement at WBN. The testing frequency requirements were then modified such that no one sample represents more than:
- 300 cubic yards for a specified strength of 2000 psi,
- 200 cubic yards for a specified strength of 3000 psi,
- 150 cubic yards for specified strengths more than 3000 psi.
For actual application, the quantities of each mix produced per shift were such that the average quantities represented by test samples were less than that specified.
- d. Aggregate - Tests are specified by ANSI N45.2.5 which appear inappropriate to certain aggregates. A carefully selected crushed limestone fine aggregate should not require testing for organic impurities. TVA required periodic reinspection of the quarry. The quarry strata and weathering effects did not change and therefore testing listed with 6-month frequency in ANSI N45.2.5 were not repeated.
3.8.3-8
WBN
- e. Water and ice - (See 1.b above) The chemical tests in CRD C 400 were repeated every 2 months, and any time a change in the water was suspected.
The strength test was repeated only when chemical tests results changed significantly.
- f. Fly ash was sampled every 3 truck loads and tested for fineness. Six samples were combined and tested for total requirements, see 1.a above.
- g. Cement - TVA accepted manufacturers' mill tests which represented no more than 400 tons. TVA made tests at greater intervals which checked manufacturers' strength test within 600 psi or duplicate tests were required.
- 3. Historical Information - TVA's concrete acceptance does not conform to ACI 318. It does conform to ACI 214. TVA requires that no more than 10% of the strength test results be below the specified strength for specified strengths equal to or greater than 3,000 psi. For lower strength concrete, 20% of the strength test results may be below the specified strength. Such concrete is used where a batch of somewhat lower strength concrete is not critical and where hydration temperature limitations are critical. ACI 318 applies the criteria that the averages of all sets of three consecutive strength test results at least equals the specified strength and that not more than 1 of 100 strengths test results will be more than 500 psi below the specified strength. If the standard deviation of the strength test results is 500 psi, the required over-strengths from the three criteria range between 640 psi and 670 psi. TVA does not believe that the three criteria produce significantly different results. ACI 318 states that acceptability is based on no strength test result being more than 500 psi below the specified strength, but its commentary and ACI 359 point out the probability that 1 test in 100 will have results outside the standard deviation and make the ACI criteria more severe.
TVA's requirement for regular compressive strength tests at 3 days and thorough evaluation requirements if the tested concrete strength deviates from the specified limits provide reasonable assurance that the use of low strength concrete in structures is effectively prevented.
- 4. Historical Information - Personnel qualifications will be maintained as required by Nuclear Quality Assurance Plan.[1]
TVA considers the applicability of ANSI N45.2.6 (Section 1.1, Scope) to be limited to those personnel performing inspection, examination, and test functions. Responsibility for examination and certification of these individuals has been established. These certifications do not correspond to the levels established in ANSI N45.2.6, except for NDE personnel who are certified in accordance with SNT-TC-1A. Construction site inspection, examination, and testing personnel are selected and assigned mechanical, electrical, instrumentation, civil, material, or welding classifications. Responsible supervisors in the respective areas perform the functions identified in Table 1 as L-III in ANSI N45.2.6. Inspection, examination, and testing personnel in the various classifications perform the functioning identified in Table 1 as L-I and L-II in ANSI N45.2.6. 3.8.3-9
WBN Bolts-Anchors set in hardened concrete were installed in accordance with comprehensive TVA specific requirements developed for the material, installation and testing of these anchors, utilizing anchor manufacturer's instructions as applicable. Welding, non-destructive examinations, heat treatment, and all field fabrication procedures used during construction were in accordance with the ASME Boiler and Pressure Vessel Code as applicable (see Item 3 below), and the American Welding Society, (AWS) "Structural Welding Code," AWS D1.1 (see Item 5 below). Nuclear Construction Issues Group documents NCIG-01 and NCIG-02 (see Section 3.8.1.2, Item 12) may be used after June 26, 1985, to evaluate weldments that were designed and fabricated to the requirements of AISC/AWS. Unless otherwise indicated in the UFSAR, the design and construction of the interior structures are based upon the appropriate sections of the following codes, standards, and specifications. Modifications to these codes, standards and specifications are made where necessary to meet the specific requirements of the structures. Where date of edition, copyright, or addendum is specified, earlier versions of the listed documents were not used. In some instances, later revisions of the listed documents were used where design safety was not compromised.
- 1. American Concrete Institute (ACI)
ACI 214-77 Recommended Practices for Evaluation of Strength Test Results of Concrete ACI 315-74 Manual of Standard Practice for Detailing Reinforced Concrete Structures ACI 359 Standard Code for Concrete Reactor Vessels and Containments (Proposed ACI-ASME Code ACI-359 (Article CC-3300) as issued for trial use April 1973. ACI 318-71 Building Code Requirements for Reinforced Concrete ACI 347-68 Recommended Practice for Concrete Formwork ACI 305-72 Recommended Practice for Hot Weather Concreting ACI 211.1-70 Recommended Practice for Selecting Proportions for Normal Weight Concrete ACI 304-73 Recommended Practice for Measuring Mixing, Transporting, and Placing Concrete
- 2. American Institute of Steel Construction (AISC) 'Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings,' adopted February 12, 1969.
- 3. American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code Sections II, III, V, VIII, and IX, 1971 Editions, as amended through summer 1972 Addenda.
- 4. American Society for Testing and Materials (ASTM), 1975 Annual Book of ASTM Standards.
3.8.3-10
WBN
- 5. American Welding Society (AWS)
"Structural Welding Code, AWS D1.1-72, with Revisions 1-73 and 2-74 except later editions may be used for prequalified joint details, base materials, and qualification of welding procedures and welders.
Visual inspection of structural welds will meet the minimum requirements of Nuclear Construction Issues Group documents NCIG-01 and NCIG-02 as specified on the design drawings or other engineering design output. See Item 14 below. AWS D12.1-61, 'Recommended Practice for Welding Reinforcing Steel, Metal Inserts, and Connections in Reinforced Concrete Connections.'
- 6. Crane Manufacturers Association of America, Inc. C.M.A.A. No. 70, Specification for Electric Overhead Traveling Cranes, 1971.
- 7. 'Uniform Building Code,' International Conference of Building Officials, Los Angeles, 1970 Edition.
- 8. Southern Standard Building Code, 1969 Edition, 1971 Revision.
- 9. 'Nuclear Reactors and Earthquakes,' USAEC Report TID-7024, August, 1963.
- 10. American Society of Civil Engineers Transactions Volume 126, Part II, Paper No.
3269, 'Wind Forces on Structures,' 1961.
- 11. Code of Federal Regulations Title 29, Chapter XVII, 'Occupational Safety and Health Standards,'Part 1910.
- 12. NRC Regulatory Guides (RG)
RG 1.12 Instrumentation for Earthquakes RG 1.31 Control of Ferrite Content in Stainless Steel Weld Metal RG 1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures RG 1.15 Testing of Reinforcing Bars for Category I Concrete Structures RG 1.55 Concrete Placement in Category I Structures
- 13. Structural Engineer Association of California, 'Recommended Lateral Force Requirements and Commentary;' 1968 Edition.
3.8.3-11
WBN
- 14. Nuclear Construction Issues Group (NCIG)
NCIG-01, Revision 2 - Visual Weld Acceptance Criteria (VWAC) for Structural Welding NCIG-02, Revision 0 - Sampling Plan for Visual Reinspection of Welds The referenced NCIG documents may be used after June 26, 1985, for weldments that were designed and fabricated to the requirements of AISC/AWS. NCIG-02, Revision 0, was used as the original basis for the Department of Energy (DOE) Weld Evaluation Project (WEP) EG&G Idaho, Incorporated, statistical assessment of TVA performed welding at WBNP. Any further sampling reinspections of structural welds subsequent to issuance of NCIG-02, Revision 2, are performed in accordance with NCIG-02, Revision 2 requirements. The applicability of the NCIG documents is specified in controlled design output documents such as drawings and construction specifications. Inspectors performing visual weld examination to the criteria of NCIG-01 are trained in the subject criteria.
- 15. TVA Reports CEB 86 Study of Long Term Concrete Strength at Sequoyah and Watts Bar Nuclear Plants CEB 86-19-C - Concrete Quality Evaluation
- 16. NRC Standard Review Plan, NUREG-0800, Rev. 2, Section 6.2.1.2, "Subcompartment Analysis".
3.8.3.3 Loads and Loading Combinations Loading combinations and allowable stresses are shown in Table 3.8.3-1. General loads are described below. Dead Loads These loads consist of the weight of the structure and equipment, plus any other permanent load contributing stress such as hydrostatic pressure. Live Loads These are movable loads such as loads which occur during servicing equipment, crane loads, and water loads due to temporary flooding of various compartments. Normal Temperature These are the straight line temperature gradients which exist through member thicknesses due to differences in operating temperatures of various compartments. LOCA Pressure These loads are time-varying pressure differentials that will result between compartments in the event of a double-ended-break of a reactor coolant pipe, as discussed in Chapter 6. They vary 3.8.3-12
WBN in magnitude depending on the location of the pipe break. During the construction permit stage, the maximum calculated differential compartment pressures were increased by 40% in accordance with NRC requirements to account for uncertainties. At the operating license stage, the design pressures equaled or exceeded the peak calculated differential pressures. Dynamic load factors were not applied in the structural analysis except for the ice condenser support floor, the walls at the end of the ice condenser compartment, and the beam over the main steam pipe in the steam generator enclosures since the vibration period of other components in the structure is small in comparison to the rate of application and duration of the pressure loads. LOCA Temperature Time-varying nonlinear temperature gradients due to a LOCA cause stresses in the restrained members of this structure. These gradients will vary depending upon time and member location in relation to the pipe break. Stresses were computed for these loadings using a TVA developed program which has the same basic assumptions as the Reinforced Concrete Chimneys Code (ACI 505-54). A typical gradient for the divider floor, is shown in Figure 3.8.3-19. Creep and Shrinkage Creep was not considered in the design of interior concrete for the reasons outlined in Section 3.8.1.4. Shrinkage effects were considered as outlined in Section 3.8.1.4. The peak hydration temperature of the concrete used in the interior structures was estimated to be approximately 130° F for summer placement with controlled placing temperatures of 65°F. From Figure 3.8.3-19 the average normal operating temperature for combination of shrinkage effects with other loads was 80°F resulting in a design temperature drop of 50°F. Under LOCA temperature gradients average temperatures exceed hydration temperatures and shrinkage stresses are relieved. Therefore, shrinkage effects are considered only with normal operating temperature gradients. Operating Basis Earthquake (OBE) Reference Section 3.7.2. Safe Shutdown Earthquake (SSE) This is the maximum postulated earthquake the plant is designed to withstand and still permit a safe shutdown. Reference Section 3.7.2. Pipe Forces These forces are the pressure jet effects that can occur due to breaks in the system's piping. They may be the jet force impinging upon the structure, or the equipment and piping anchorage forces as the result of such a jet. The static equivalent of the major equipment anchorage loadings were furnished by Westinghouse Corporation, the support designer. Jet forces from postulated pipe ruptures, both longitudinal and transverse, were assumed to load the interior concrete structure. F = (1.2pA)k where: F = Force on structure, lbs p = Pressure, psi 3.8.3-13
WBN A = Inside cross sectional area of pipe, in2 k = Load factor The effects of jet force was taken in combination with the uniform compartment differential pressure. A minimum load factor k = 1.3 was used with both conditions based on localized yielding of the structural member and a ductility factor of 3. The only jet force in the compartment above the reactor cavity is from the control rod opening. This is due to a pressure of 2250 Psi. The reactor coolant pipe will not produce a jet force in this area. Missiles The systems located inside the reactor containment have been examined to identify and classify potential missiles. The basic approach is to assure design adequacy against generation of missiles, rather than allow missile formation and try to contain their effects. Reference Section 3.5. Ice Condenser Loads and Loading Combinations The ice bed structure shall be designed to meet the loads described below within the behavior criteria limits presented in Section 3.8.3.5 of these criteria. The following load combinations are defined for design purposes:
- 1. Dead Load + Operating Basis Earthquake Loads (D + OBE)*.
- 2. Dead Load + Accident induced loads (D + DBA).
- 3. Dead Load + Safe Shutdown Earthquake (D + SSE).
- 4. Dead Load + Safe Shutdown Earthquake + Accident induced loads (D + SSE + DBA).
- Includes thermal induced load and D+L The loads are defined as follows:
Dead load (D) - Weight of structural steel and full ice bed at the maximum ice load specified. Live Load (L) - Live load includes any erection and maintenance loads, and loads during the filling and weighing operation. Thermal Induced Load - Includes those loads resulting from differential thermal expansion during operation plus any loads induced by the cooling of ice containment structure from an assumed ambient temperature at the time of installation. 3.8.3-14
WBN Accident Fluid Dynamic and Pressure Loads (DBA) - Accident pressure load includes those loads induced by any pressure differential drag loads across the ice beds, and loads due to change in momentum. Operating Basis Earthquake (OBE) - As previously defined. Safe Shutdown Earthquake (SSE) - As previously defined. 3.8.3.4 Design and Analysis Procedures 3.8.3.4.1 General Each component of the interior concrete structure was considered individually. Its boundary conditions and degrees of fixity were established by comparative stiffness; loads were applied, and moments, shears, and direct loads determined by either moment distribution or finite element methods of analysis. Reinforcing steel was proportioned for the component sections using the allowable stresses given in Table 3.8.3-1, the provisions of the ACI 318-71 Building Code and the proposed Standard Code for Concrete Reactor Vessels and Containments, ACI-ASME (ACI-359) Code, as issued for trial use, April 1973. During the construction stage, a factor of 1.4 was applied to the design pressures resulting from LOCA. The structure was designed using the 40% margin and the recommendations of the ACI-ASME Joint Committee contained in Proposed Standard Code for Concrete Reactor Vessels and Containment. The results are tabulated in Table 3.8.3-2. NRC Standard Review Plan, NUREG-0800, Rev. 2, Section 6.2.1.2, Section II.B.5, permits reduction of design pressure so that the peak calculated differential pressure does not exceed the design pressure. This reduction in design pressure was utilized in the review of the concrete strength evaluation.[2] A completely independent design was performed on all portions of the divider barrier. Procedures used in this design and analysis are discussed in Sections 3.8.3.4.3 through 3.8.3.4.13. 3.8.3.4.2 Structural Fill Slab on Containment Floor The fill slab is designed to span between walls with hydrostatic uplift pressure on 100% of its bottom face from water surface at Elevation 710. Loads from the steam generators and reactor coolant pumps are transferred directly into the base mat by continuous steel connections through the liner plate. As the base mat deflects under load, the fill slab deflects with it and is designed for these deflections. Analysis was made using the effects of crane wall uplift and rotation as well as the uplift of the support columns for the Elevation 716.0 floor. The fill slab is also designed for loads imposed upon it by the reactor coolant pipe crossover supports. Classical deflection formulas, as well as the computer code ICES STRUDL II, were used to determine moments, shears, and reactions. The finite element method of analysis was used to determine direct stresses in shear keys. 3.8.3-15
WBN 3.8.3.4.3 Reactor Cavity Wall In the event of a circumferential split of a reactor coolant pipe at a reactor vessel nozzle, a nonsymmetric pressure occurs in the inspection cavity region. The pressures resulting from this break, a maximum of 148 psi, are applied to the design of the 8-foot, 6-inch thick and 4-foot, 3-inch thick reactor cavity wall. Large horizontal restraint forces from the steam generators and reactor coolant pumps of approximately 2,200 kips will be transmitted to the 4-foot, 3-inch thick section of wall. A linear temperature gradient will occur during reactor operation. In the event of a LOCA the reactor is shut down and time-varying nonlinear, gradients similar to those in Figure 3.8.3-19 will occur in the wall. All temperature gradient cases are considered in the design up to a time of 48 hours following a LOCA. Radiation generated heat on the structures is considered only for the primary shield immediately next to the reactor vessel. There the radiation generated heat is obtained as a function of position of the reactor core with respect to the structure and the temperature distribution is calculated. The effect of the temperature on the structure is then evaluated. The average temperature of the wall during reactor operation exceeds hydration temperatures during construction. Therefore, tensile stress from hydration temperature considerations will be less than the stress induced by the temperature gradient. Long-time creep relaxation can be expected to substantially reduce temperature stresses; however, such a reduction was not utilized since the effective operating temperature differential across the wall of the reactor cavity was only 35°F. The 81/2-foot-thick portion of the wall is basically a thick cylinder. Formulas for stresses in thick cylinders were used to determine the ring moments and tensile stresses induced by the pressure loading and thermal gradients. The 81/2-foot-thick portion of the wall was also analyzed as fixed at its base, Elevation 702.78, and an analysis for vertical moments and forces was made for LOCA, reactor support loads and thermal effects. The 4-foot structural portion from the top of the 81/2-foot-thick ring to the top of the reactor continues in the same shape and configuration as the compartment above the reactor. This portion of the wall was modeled as a continuation of the compartment above the reactor using the computer program SAP IV (1973). The thin plate and shell element of this program was utilized. The wall was considered fixed at its base by the 8-1/2-foot-thick ring. The large openings for the reactor coolant system pipes occur in this portion of the wall and they were taken into account in the computer model. The axisymmetric pressure due to a reactor coolant pipe break was applied to the wall as well as the large concentrated forces from the steam generator and reactor coolant pump anchorages. Moments, shears, axial loads, and displacements were obtained from the computer analysis and reinforcing selection was made from these results. Independent Design - Historical Information The analysis of the reactor cavity was approached from the standpoint that a dynamic and inelastic type analysis might be required. However, comparison of the natural frequencies of the structure and the shock spectra of the pressure transients as forcing functions indicated that dynamic amplification was negligible. The results of static stress analyses showed that the stresses in the concrete were generally less than the cracking stress of concrete except in the immediate area of the break. Consequently, only a static elastic analysis was necessary. Analyses were performed for the pressure loads at specific times combined with support loads from reactor coolant loop and reactor pressure vessel. Thermal stresses resulting from this LOCA were also combined directly with pressure-induced stress. The structural model used in the independent review was a three-dimensional assemblage of intersecting walls simulated by multiple layers of solid isoparametric finite elements. The 3.8.3-16
WBN general purpose computer program ANSYS, a well documented and widely used program, was used to calculate stress intensities. The reactor cavity structure was considered fixed at the intersection of the reactor cavity columns and operating deck and at the base. A detailed analysis to determine unit stresses in the concrete and reinforcing steel was based on a cracked section and evaluated using the allowable stresses given in Table 3.8.3-1. 3.8.3.4.4 Compartment Above Reactor This compartment, which has a design internal pressure of 32 psi, will also be subjected to water pressure when the refueling canal is full of water and refueling of the reactor is taking place. Wave effects of the water during earthquake are taken into account. This compartment was analyzed in conjunction with the refueling canal walls using the SAP IV (1973) computer program. The reactor cavity wall, as well as the compartment above the reactor, are post-tensioned during construction. Tendons are located at approximately 45° increments around the wall at seven locations. These post-tensioning tendons are stressed as soon as the concrete in the walls has reached a proper strength. This stressing operation puts a 1,000 kips compressive load in the concrete walls at each of the tendon locations. This compressive force is taken into consideration in the reinforcing design of both the reactor cavity and the compartment above the reactor walls. Independent Design - Historical Information This compartment was analyzed in conjunction with the reactor cavity wall as described in Section 3.8.3.4.3. The 4-foot-thick compartment wall was designed for, a) a break at the reactor vessel nozzle in the lower cavity, b) a break in a primary coolant loop outside the reactor cavity structure along with associated support loads, c) a hydrostatic load applied to the interior face during refueling, and d) CRD mechanism restraint loads. The compressive stress due to post-tensioning of the walls was considered in the formulation of the finite element model. 3.8.3.4.5 Seals Between Upper and Lower Compartments The design of the seals was by TVA without the use of a computer program. The flexible coated fabric part of the seal was considered as a thin-wall half cylinder as the fabric width was sized to form an approximate semicircle when subject to internal pressure. With the semicircle, there is adequate slack in the seals to provide for relative movement between the attaching surfaces during all conditions without damage to the seals. Earthquakes are the only natural environmental conditions which apply to the seals. The seals, being inside the containment vessel are protected from floods, wind, tornadoes, snow and ice. The seals are not in the area affected by missiles and therefore were not designed for missiles. 3.8.3-17
WBN 3.8.3.4.6 Refueling Canal Walls and Floor (Divider Barrier) Primary Design The canal walls and slab are designed to take the gravity and earthquake forces from the upper and lower internals storage stand. The face of the walls inside the lower compartment will be subject to a maximum LOCA pressure of 24 psi and localized jet forces due to a LOCA. The walls are subject to concentrated forces and moments from the reactor coolant pump restraints. The walls are subject to an uplift condition due to LOCA pressure acting on the divider barrier slab at Elevation 754.13. The canal walls and slab are designed for the water pressure in the canal during refueling operations. The seismic effect on this water was also considered. The walls of the refueling canal and the compartment above the reactor were analyzed as a unit consisting of both straight and curved sections of walls. These walls were analytically modeled using the SAP IV (1973) finite element computer program. Shell curved-rectangular elements were used in the mesh assembly and spring constants were used to represent the stiffnesses of walls and slabs framing into the canal walls. Spring constants were used at the intersection of the canal walls with the crane wall, operating deck slab, and the canal floor slab. The refueling canal slab was analyzed using the STRUDL finite element computer program. Both the rectangular and triangular flat plate element were used in the analysis. Independent Design - Historical Information
- 1. Refueling Canal Floor The refueling canal floor slab was modeled and analyzed utilizing the SAP IV finite element program. The floor, due to its irregular shape, was analyzed in two sections.
The larger area of the floor inside the crane wall radius was analyzed utilizing a finite element plate bending model. Boundary conditions were input reflecting the relative stiffnesses of the supporting walls. The slab was fixed at the crane wall boundary. A smaller segment beyond the crane wall support was analyzed using the conservative "strip" method of analysis. SAP IV is a general structural analysis program for the static and dynamic analysis of linearly elastic structures. This program was developed and published by Bathe, Wilson, and Peterson, The College of Engineering, University of California at Berkeley, in June 1973, and has seen extensive usage since that time. The plate element used in the analysis is a quadrilateral of arbitrary geometry formed from four compatible triangles. A constant strain triangle was used to represent membrane stresses and the LCCT9 element was incorporated to represent bending behavior. The design loads consisted of base plate forces from the upper and lower reactor internals, and fluid pressure forces from the flooded state during refueling. Earthquake and thermal gradient loading was also considered. Factored and unfactored load cases were checked and the reinforcing was sized for maximum stresses using the criteria of Table 3.8.3-1. 3.8.3-18
WBN
- 2. Refueling Canal Walls These walls were analyzed in conjunction with the reactor cavity walls as described in Section 3.8.3.4.3. Primary loads considered were, a) a break in a main steam line outside the reactor cavity with a resulting jet force and associated reactor coolant system support loads, b) a hydrostatic load during refueling with associated upper and lower internal support stand loads, and c) the effects of a main steam line break inside the reactor cavity. The model employed is described in the aforementioned Section.
3.8.3.4.7 Crane Wall Wall Below Operating Deck In the lower compartment the crane wall is subject to jet forces due to a possible break in the reactor coolant or main steam piping. The largest of these postulated jet forces is 2,650 kips occurring at a crossover leg between a steam generator and a reactor coolant pump. In this same area, the crane wall is subject to a large missile impingement load. Other areas of the crane wall in the lower compartment are exposed to uniform pressure differentials of approximately 23 psi. The steam generators and reactor coolant pumps are braced laterally with restraints anchored into the crane wall. These restraints impose large concentrated loads on the wall. The largest of these loads is approximately 2,300 kips. Crane wall temperature gradients, before and after a LOCA, were investigated. At several elevations in the crane wall, maximum and minimum vertical loads were computed using results from the "Dynamic Earthquake Analysis of the Interior Concrete Structure, prepared by TVA. In addition, various parts of the crane wall were designed to handle concentrated loads, 100 to 300 kips, resulting from breaks in small piping systems. The crane wall was analyzed by isolating areas spanning between slabs and cross walls. Moments, shears, and axial forces were calculated using the STRUDL finite element program. Fixed-end moments were distributed between adjacent sections of wall using conventional distribution methods. Columns between ice condenser doors are subjected to moments and forces distributed from the ice condenser floor and divider barrier floor, as well as moments, shears, and axial forces from the wall sections above and below the columns. The columns were designed for these moments, shears, and axial forces plus earthquake loads. The columns in the vicinity of the steam generator and pressurizer compartments are greatly influenced by the lateral restraints from the aforementioned equipment, which is anchored in the crane wall immediately at the top of the columns. Personnel Access Doors in the Crane Wall Main structural members of the doors were considered as simple beams. Energy absorbing members were considered as collapsible members. Members of the embedded frames were considered as being rigidly supported by concrete. Loads from the embedded frames are transferred to the concrete by embedded anchors. 3.8.3-19
WBN Design of the doors and embedded frames was by TVA without the use of a computer program. Design of collapsible members on the doors was based on tests made by Oak Ridge National Laboratory. Results of these tests are recorded in their publication titled Structural Analysis of Shipping Casks, Volume 9, "Energy Absorption Capabilities of Plastically Deformed Struts Under Specified Impact Loading Conditions." Collapsible members were sized to limit loads transmitted to the embedded frame to 13,000 pounds per linear inch. The doors were designed to function during normal conditions, earthquakes, and pipe rupture accidents. Earthquakes are the only natural environmental condition which applies to the doors. The doors are protected from flood, wind, tornadoes, ice and snow, since they are located inside the containment vessels. The doors will be closed any time reactor containment is required, except when a workman is passing through the access. When containment is not required, the doors are not required to seal or to retain their integrity. Since the doors are left open only when containment is not required, seismic qualifications of the doors in the open position is not required. Earthquake loads used in designing the doors were from accelerations determined for the crane wall at the horizontal centerline of each door by dynamic analysis of the Reactor Building for an OBE and an SSE. These acceleration loads were used as static loads since the doors are firmly secured to the wall when closed. Doors were reanalyzed using Set "B" ARS spectrum. Some air leakage may occur at the periphery of the doors during earthquakes, but this leakage will not exceed the permissible leakage of 30 square inches per door. Under normal conditions, seals on the doors will have a life of not less than 10 years, and the other parts of the doors will have a life of not less than 40 years. Some air leakage may occur at the periphery of the doors, but this leakage will not exceed the permissible leakage area for normal operation of 10 square inches per door. Wall Above the Operating Deck (Divider Barrier) Primary Design Under accident conditions the crane wall above the operating deck is designed for maximum pressure differentials between the ice compartments outside the crane wall and the steam generator and pressurizer compartments. It is also designed for the loads imposed on the wall by the lattice frame anchorages of the ice condenser. Maximum and minimum vertical loads imposed by the earthquake analysis are combined with these loads for the maximum stress conditions. The end walls of the ice condenser and the spacing of the steam generator and pressurizer walls stiffen the crane wall to such an extent that it essentially spans horizontally between these supporting walls. The stud loadings on the wall from the lattice frame may either add to or subtract from the pressure loading depending on whether the maximum pressure is inside the crane wall or in the ice compartment. The portion of the crane wall behind the steam generator and pressurizer compartments is subject to jet impingement loads. 3.8.3-20
WBN The steam generator and pressurizer upper lateral restraints are anchored in the crane wall and exert large forces on it. The effect of pressure on the top slab of the pressurizer and steam generator compartments causing an uplift on the crane wall was considered. The STRUDL II frame program, STRUDL II finite element program, and the SAP IV (1973) finite element program are the principal computer programs used in the analysis of the upper crane wall above the operating deck slab. The upper restraint of the steam generators is designed such that the crane wall only receives load from a steam line break. Seismic restraints in the radial direction are transmitted through hydraulic snubbers to the floor of the operating deck. In the other direction they are transmitted to the walls of the steam generator compartment. The two 720-kip steam generator restraint loads on the crane wall are assumed to occur coincidentally with the maximum pressure differential in the steam generator compartment. During construction, a 36-foot-wide opening, used for moving major equipment into the building, was left in the crane wall at approximately the 90° azimuth. This opening began at Elevation 756.63 and extended 46 feet high. This leaves a 3-foot-wide, 17-foot-deep curved beam spanning a 371/2-foot arc over the opening. This beam and the permanent beam over the refueling canal are designed to carry the construction loads of the polar crane, approximately 1200 kips maximum while installing major equipment. The permanent beam is also designed to take the reactions from the cantilevered beams supporting the ice condenser bridge crane. The analysis of these beams was made using the STRUDL computer program. The top of the crane wall is designed to withstand a force in the radial direction due to the polar crane bumping into it as a result of seismic action. This seismic force from the polar crane is considered to act at any point on the circumference of the wall and approximately 2 feet below the top of the wall. Independent Design - Historical Information The crane wall resists two general types of loads. They are, 1) localized forces from equipment supports, pressure forces, structure discontinuities, etc., and 2) forces from gross structure motions of the interior concrete structure induced by the design earthquakes and the design basis accident (DBA). Calculations of gross forces in the interior concrete structure due to a design earthquake are described in Section 3.7.2.1.1. The lumped mass cantilever beam model used in the seismic analyses was loaded by a time-dependent forcing function representing the non-axisymmetric pressure loads from a DBA. The analysis used modal superposition and determined the total responses in the time domain. The results of the analysis consisted of gross overturning moments and shears and accelerations, deflections, and acceleration response spectra at various elevations. Portions of the crane wall subjected to very isolated forces were isolated and designed as substructures. Boundary conditions were always chosen to give conservative results. The forces at the junction of crane wall and other divider barrier components such as the operating deck, ice condenser floor, and end walls were included in the design. 3.8.3-21
WBN-1 The crane wall at the ice condenser inlet doors between Elevations 746.42 to 753.63 were designed as beam columns. The columns are subjected to both localized and gross motion forces. Localized loads resulted from the steam generator support loads, pressure forces, and interactions from the operating deck and the ice condenser floor. The column design was verified by both working stress and ultimate strength methods. The portion of the crane wall within the steam generator compartment was analyzed as part of the steam generator enclosure. This portion was modeled using the MARC-CDC nonlinear finite element computer program. The element utilized was the 20-node isoparametric solid. The loadings on the wall consisted of reaction loads from the steam generator supports near the crane wall columns and pressure loads from postulated breaks in the main steam line. The crane wall columns were designed to resist maximum moments, shears, and axial loads due to the local forces on the crane wall as well as forces due to gross motions of the interior concrete. The crane wall segment of the steam generator compartment model was extended past the juncture of the enclosure and the crane wall to minimize boundary effects on the solution. Boundary conditions assumptions were selected to provide conservative stresses. The crane wall within the pressurizer compartment was designed by a similar method. 3.8.3.4.8 Steam Generator Compartments (Divider Barrier) These compartments are designed to resist a maximum of 38 psi differential pressure on the wall common with the upper compartment and 26 psi differential pressure on the center wall that would result following a main steam pipe break inside any single compartment. The center wall is also designed for the effect of a 1160-kip jet force that would result from a main steam pipe break. Also accounted for are thermal effects accompanying a pipe break (see Figure 3.8.3-19). The compartments span mainly in the horizontal direction resulting in tensile stresses and horizontal moments in the walls near the center of their height. Close to the ends of the compartments, discontinuity stresses result in the vertical direction, similar to those of a flat head cylinder. The STRUDL frame program was used to find the maximum horizontal forces in the walls by modeling a vertical 1-foot height of walls including a 113-degree sector of crane wall. Short chord lengths were used to represent curved sections of walls. Manual calculations were done at the top and bottom of the wall which is common to the upper compartment to investigate the effects of the slabs restraining the wall. In addition, the flat plate finite element STRUDL computer program was used to analyze the center wall for moments and shears in both directions. The top slab was analyzed using stiffened members in the flat plate finite element STRUDL program. The inverted "tee"-shaped beam which stiffens the top slab and which is located at the top of the center wall was analyzed for the dynamic effects of a main steam pipe breaking and striking the flange of the beam. Reanalysis Due to Unit 1 Steam Generator Replacement The modified configuration of the Unit 1 steam generator compartment was analyzed for design loads using a 3D finite element ANSYS model. Although the roof slab remains the focus of the evaluation, the model included five components - the 3 feet thick roof slab, entire steam generator compartment wall, center wall, 180° sector of the crane wall and the whip-restraint beam; to obtain an accurate representation of the system. The material properties used in the model for the concrete were consistent with those used in the original analysis. The concrete strength used in the roof evaluation is the in-place compressive strength of the steam generator compartment roof concrete at 90 days. 3.8.3-22
WBN-1 The modified steam generator compartment roof was analyzed for the following design loads: dead load, live load, design pressure differential from a DBA (main steam pipe break), operating and accident temperature effects, seismic effects (OBE and SSE), and pipe thrust load on the whip-restraint beam from a broken main steam pipe. The modified steam generator compartment roof was evaluated to the load combinations and allowable stresses tabulated in Table 3.8.3-1. Independent Design - Historical Information
- 1. Roof Slab The steam generator enclosure roof slab was analyzed as a thick plate using the three-dimensional 20-node isoparametric solid element available in the MARC-CDC finite element program. The T-beam stiffener attached to the inside of the slab was included in this finite element model.
The MARC-CDC finite element program is based on research work carried out by Professor Pedro V. Marcal of Brown University and colleagues at the University of London. In 1969 the program was released commercially by the Marc Analysis Research Corporation. The Control Data Corporation has recently documented the program and offered it for general usage through their computer system. A sample problem was run to verify the results obtained from the program. The boundary conditions of the roof slab were approximated by modeling the actual stiffness of the steam generator enclosure wall and crane wall. The design loads considered originated from a postulated rupture in the main steam line in a steam generator compartment. Differential pressures, jet force and pipe reaction resulting from the postulated break were checked for factored and unfactored load cases. The T-beam beneath the roof slab was designed to resist pipe whip forces from a postulated guillotine break in the main steam line. Moments, shears, and torsional stresses were checked to ensure adequate reinforcing utilizing the stress criteria of Table 3.8.3-1.
- 2. Enclosure Walls and Separation Wall The steam generator enclosure wall, separator wall, and a segment of the crane wall were modeled utilizing the three-dimensional 20-node isoparametric element available in the MARC-CDC nonlinear finite element program.
Spring constants simulating the crane wall columns were obtained by analyzing a small portion of the crane wall. The walls of the steam generator enclosure were then modeled to include appropriate boundary conditions at the intersection of the operating deck, upper crane wall, and the crane wall columns. The loadings considered were of two basic types: a) a pressure load obtained from a hypothetical DBA, and b) snubber and embedment plate loads caused by combined LOCA and seismic action. Live loads, dead loads, seismic loads, and thermal loads were also considered. Flexural, axial, shear, and torsional stress levels resulting from the factored and unfactored loading combinations were evaluated using the allowable stress criteria of Table 3.8.3-1. Reinforcement was selected based on these criteria. 3.8.3-23
WBN 3.8.3.4.9 Pressurizer Compartment (Divider Barrier) Primary Design The compartment is designed to resist a 50 psi differential pressure. Methods of analysis were similar to those of the steam generator compartments. Independent Design - Historical Information The enclosure wall and the adjacent crane wall were modeled using solid isoparametric finite elements. The computer program utilized was the previously documented SAP IV. The crane wall model was extended beyond its juncture with the enclosure wall to minimize boundary effects. Boundary conditions at the intersection of the enclosure wall with adjacent elements of the interior concrete structure were chosen to provide conservative stress results. The roof slab was modeled as a thick plate using the ANSYS computer program, a widely used and previously documented program. Boundary conditions at the junction with the crane and enclosure walls were chosen to provide maximum stress levels. With these stress levels, a detailed analysis to determine unit stresses in the concrete and reinforcing steel was based on a cracked section and evaluated using the allowable stresses given in Table 3.8-3-1. 3.8.3.4.10 Operating Deck at Elevation 756.63 (Divider Barrier) Primary Design The floor is designed to carry a 24 psi upward pressure and thermal effects due to a LOCA plus the jet pressure of 340 psi acting over a local area. Upward loads from the missile shield are taken by this floor around the reactor cavity where the shield is bolted down. This floor is designed for a 1,000-psf live load. This loading suffices for several concentrated loads from the reactor head set down which occur during refueling and periodic maintenance of the equipment. During construction, the floor has no edge support at the steam generator and pressurizer compartment walls, since the first lifts of these walls are carried by the floor. This special construction condition was examined separately using a 300-psf design live load in addition to the wet weight of the first 45-foot pour of walls. The floor analyses were made using the STRUDL finite element program for flat plates utilizing both rectangular and triangular elements to assemble the irregular shape. Anchorage for the upper steam generator restraints is provided in this floor. These anchorage points have approximately a 5,500 kip design force due to a LOCA combined with SSE. This force is horizontal and applied at points where openings create horizontal single span beams. These beams were analyzed using manual methods. 3.8.3-24
WBN Independent Design - Historical Information The operating deck was analyzed using four finite element models to represent this irregularly shaped slab. Two models were used to analyze the larger segment from 0° to 180° while two additional models were used to analyze the segments adjacent to the refueling canal walls. Loads normal to the surface and inplane loads act on the operating deck. Stresses from concentrated snubber (inplane) loads were determined from a plane stress finite element analysis. These snubbers act as the lateral supports for the steam generators. In addition to the snubber loads, pressure loads and jet forces from postulated pipe ruptures acting normal to the operating deck were determined from a plate bending finite element model. Dead, live, seismic, pipe support, and construction loads were added to the differential pressure and snubber forces to complete the factored and unfactored load cases. The SAP IV plate bending and plane stress elements were used to calculate moments, shears, reactions, and inplane stresses. Boundary conditions at the crane wall, steam generator enclosure, pressurizer enclosure, reactor cavity, and refueling canal wall junctions were chosen to give the most conservative results. Using the calculated moment, shears, and axial forces, stresses in the reinforcement for an assumed cracked section were checked against the stress criteria of Table 3.8.3-1. 3.8.3.4.11 Ice Condenser Support Floor - Elevation 744.5 (Divider Barrier) Primary Design The finite element program SAP IV (1973) was primarily used in the analysis and design of the Elevation 744.5 floor. The outer circumferential beam was represented along with the floor by using a combined flat plate and grid member system. The supporting columns were modeled using spring constants for both rotation and deflection. Shear and moment values were obtained from the computer program at the crane wall, ice condenser end walls, and supporting columns. Reinforcing selections were made from these results. Independent Design - Historical Information The ice condenser floor was analyzed as a series of circumferentially beam-stiffened curved slabs with a continuous support along the inner radius and ends. These slabs were supported along the outer radius by flexible columns. The analysis was made utilizing SAP IV, a previously documented computer program. Models were generated for segments of the ice condenser floor at 0°, 90°, 44°, 180°, and 230°. Boundary conditions between the segments were input to provide maximum forces at the boundary and midspan. The various segments of the ice condenser floor were dynamically analyzed to determine natural frequencies. Next, time varying differential pressure loads were used to calculate dynamic load factors associated with the natural frequencies of each segment. The dynamic load factors obtained were then applied to the maximum differential pressures in addition to the factors for the ice condenser structure support loads. 3.8.3-25
WBN The loadings evaluated included concentrated forces and moments from the ice condenser support system, jet forces resulting from change in momentum of the steam flow, and LOCA induced pressure differentials. The ice condenser support system loads resulted from a combination of seismic accelerations and drag on the ice baskets by the channeled gases. Normal and factored load cases were examined to determine maximum moments and shears. Stresses in the concrete and reinforcement were evaluated against the criteria of Table 3.8.3-1. 3.8.3.4.12 Ice Condenser Analysis, meeting the criteria presented in Section 3.8.3.5, has been done on the basis of elastic system and component analyses. Limit load analysis was used as an alternate to the elastic analysis. Limit loads are defined using limit analysis by calculating the lower bound of the collapse load of the structure. Load factors are applied to the defined design basis loads and compared to the limit loads. The load factors determined for design basis load are used to provide margins of safety of the structure against collapse. A load factor of 1.43 was used when considering the mechanical loads due to dead weight and OBE. A load factor of 1.3 was used for (D + SSE) and (D + DBA). The material was assumed to behave in an elastic-perfectly-plastic manner where strain-hardening effects are neglected. The minimum specified yield strength was used. Mechanical plus thermal induced load combination and fatigue was analyzed in an elastic basis and satisfy the limits of Section 3.8.3.5. The stress analyses and results are described in Sections 3.7 and 6.7. Experimental or Test Verification of Design In lieu of analysis, experimental verification of design using actual or simulated load conditions was used. In testing, account was taken of size effect and dimensional tolerances (similitude relationships) which exist between the actual component and the test models, to assure that the loads obtained from the test are a conservative representation of the load carrying capability of the actual component under postulated loading. The load factors associated with such verification are: 1.87 for D + SSE, 1.43 for D + DBA or D + SSE, and 1.3 for (D + SSE) or 1.3 for (D + DBA). A single test sample is permitted but in such cases test results were derated by 10%. Otherwise, at least three samples were tested and the design was based on the minimum load carrying capability. Additional analysis results are found in Section 6.7. 3.8.3-26
WBN 3.8.3.4.13 Penetrations Through the Divider Barrier Canal Gate Primary Design The canal gate sections are designed to span as simply supported beams across the refueling canal; a clear span of some 19 feet. Hand calculations using conventional methods were used for this design. The canal gate was designed to withstand a 39-psi pressure differential between the compartment above the reactor and the upper compartment due to a LOCA. The effect of seismic action on the canal gate sections is considered as well as the effect of maximum temperature differential across the gate. Independent Design - Historical Information The canal gate was designed as a simply supported plate spanning between the two refueling canal walls. The canal gate is required to maintain integrity between the upper and lower compartments during a LOCA and was designed for the maximum probable differential pressure. The effects of seismic and thermal action were evaluated. Moments and shears were calculated using conventional hand methods. The evaluation of concrete and reinforcing steel stresses was based on a cracked section and the allowable stresses of Table 3.8.3-1. Control Rod Drive (CRD) Missile Shield Primary Design The CRD missile shield sections are designed to span as simply supported slabs across the compartment above the reactor. The slabs are held down at the ends by anchor bolts embedded in the operating deck slab. The missile shield is designed to withstand a maximum differential pressure of 39 psi between the compartment above the reactor and the upper compartment due to a LOCA. The missile shield is subject to loading from the CRD mechanism as a missile. An accompanying jet force due to pressure escaping through the head of the reactor is also considered. The slabs are investigated for the maximum penetration resulting from the missile effects of the control rod drive shaft. The underside of the slab is faced with a 1-inch-thick steel plate to aid in resisting missile penetration. The penetration depths are calculated by use of the Petry formula and a formula by C. V. Moore, "The Design of Barricades for Hazardous Pressure Systems," Nuclear Engineering and Design 5 (1967), 81-97, North-Holland Publishing Company, Amsterdam. The calculated penetration depth is 2.2 inches into the 3-foot, 6-inch thick slab. The effect of maximum temperature differential across the missile shield is also considered in the design. 3.8.3-27
WBN Independent Design - Historical Information The missile shield sections were analyzed as simply supported slabs spanning the compartment above the reactor. This shield must resist the maximum probable differential pressure from a LOCA to maintain integrity between the lower and upper compartments. Additionally, it must resist certain missiles from the control rod drive mechanism. Penetration into the steel plate and concrete were calculated and equivalent static loads for the impacting missiles were calculated and evaluated. Thermal stresses resulting from temperature differentials between lower and upper volumes were considered in the stress evaluation. Concrete and reinforcing steel stresses were determined considering a cracked section and the stress allowables of Table 3.8-3-1. Reactor Coolant Pump Access and Lower Compartment Access Hatches Primary Design The reactor coolant pump access and lower compartment access slabs are designed to span simply supported between anchor bolt. The slabs are designed for both downward and upward loads acting on them. The downward loads are dead load and a 1,000-psf live load. For upward loads, the slabs are designed to carry a 24-psi differential pressure between the lower and upper compartments due to a LOCA. A jet impingement loading associated with this LOCA is also considered. The effect of maximum temperature differential as well as seismic effects on the slabs are accounted for in the design. Independent Design - Historical Information The reactor coolant pump access hatch and the lower compartment equipment access hatch were analyzed and designed as simply supported circular and rectangular plates. Maximum moments and shear forces were obtained from a plate bending analysis. Dead, live, seismic, and thermal loads were combined with differential pressures and jet forces due to a postulated LOCA to give controlling factored and unfactored load cases. Shear stresses at the periphery of the hatch openings in the operating deck and stress levels in the perimeter anchor bolts were checked to ensure compliance with criteria of Table 3.8.3-1. Equipment Access Hatch The hatch cover is designed to span as a simply supported beam between the anchor bolt assemblies with the anchor bolts designed to withstand a load at least 5% greater than that calculated for the end reactions resulting from the actual load on the hatch. The allowable stresses are given in Table 3.8.3-6. Escape Hatch Structural components of the hatch have been designed such that, the allowable stresses given in Table 3.8.3-7 will not be exceeded. 3.8.3-28
WBN Air Return Duct Penetrations The controlling design condition is Design Basis LOCA Pressure - Dead Load + Safe Shutdown Earthquake Loads. Maximum differential temperature is considered in the design, but does not occur coincidentally with a jet force or the maximum differential Design Basis LOCA Pressure. Calculated and allowable stresses are given in Table 3.8.3-8. 3.8.3.5 Structural Acceptance Criteria 3.8.3.5.1 General Structure Calculated and allowable stresses for the principal concrete structure are listed in Table 3.8.3-2. Locations of these areas are shown referenced to Figures 3.8.1-1, 3.8.3-6 and 3.8.3-7. The working stress design method was used for finding stresses. 3.8.3.5.2 Structural Fill Slab on Containment Floor During the original design (construction permit) phase with 40% added to LOCA pressure, the controlling combination was "Abnormal/Severe Environmental." See Table 3.8.3-2. Shear transfer through the fill slab is discussed in Section 3.8.3.1.2. 3.8.3.5.3 Reactor Cavity Wall and Compartment Above Reactor Loading combinations 1 through 7 in Table 3.8.3-1 were considered in the design. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combinations are "Abnormal" and "Abnormal/Severe Environmental." See Table 3.8.3-2. The 4-foot, 3-inch structural thickness provided adequate depth for limiting peripheral shear stresses due to anchorage loads on the steam generator and reactor coolant pump restraints. Earthquake shears for the interior structures were distributed to the various walls in proportion to their rigidity. This term is defined as 1/ where delta is the deflection due to a unit force, and it includes deflection due to bending and shear. The procedure is outlined in "Analysis of Small Reinforced Concrete Buildings" by the Portland Cement Association. The columns at the top of the compartment above the reactor wall were designed for the effect of earthquake shears by proportioning shear to them based upon their rigidity as previously described. 3.8.3.5.4 Refueling Canal Walls and Floor Loading combinations 1 through 7 in Table 3.8.3-1 were examined. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combinations are "Abnormal" and "Abnormal/Severe Environmental." See Table 3.8.3-2. 3.8.3-29
WBN 3.8.3.5.5 Crane Wall The crane wall was analyzed for loading combinations 1 through 7 in Table 3.8.3-1. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combination is "Abnormal/Severe Environmental." See Table 3.8.3-2. Earthquake shears were calculated utilizing the procedure discussed in Section 3.8.3.5.3. Shear reinforcement was required in many areas of the crane wall for radial shears generated by jet impingement loading, as well as pipe and equipment restraint reactions, missile impinge-ment loading and pressure due to LOCA. 3.8.3.5.6 Steam Generator and Pressurizer Compartment Loading combinations 1 through 7 in Table 3.8.3-1 were examined. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combination is "Abnormal" for the pressurizer compartment and "Abnormal/Severe Environmental" and "Abnormal/ Extreme Environmental" for the steam generator compartment. See Table 3.8.3-2. 3.8.3.5.7 Operating Deck at Elevation 756.63 Loading combinations 1 through 7 in Table 3.8.3-1 were examined for the floor. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combination is "Abnormal/Severe Environmental." See Table 3.8.3-2. 3.8.3.5.8 Ice Condenser Support Floor - Elevation 744.5 Loading combinations 1 through 7 in Table 3.8.3-1 were examined in the design of the floor. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combination is "Abnormal." See Table 3.8.3-2. 3.8.3.5.9 Penetrations Through the Divider Barrier Canal Gate and Control Rod Drive (CRD) Missile Shield Loading combinations 1 through 7 in Table 3.8.3-1 were examined. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combination is "Abnormal/Severe Environmental." See Table 3.8.3-2. Reactor Coolant Pump and Lower Compartment Access Hatches Loading combinations 1 through 7 in Table 3.8.3-1 were examined. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combination is "Abnormal/Severe Environmental." 3.8.3-30
WBN Escape Hatch Loading combinations 1 through 7 in Table 3.8.3-1 were examined. During the original design (construction permit) phase with calculated values of LOCA pressure load increased by 40%, the controlling load combination is "Abnormal/Severe Environmental." 3.8.3.5.10 Personnel Access Doors in Crane Wall Allowable stresses for non-collapsible members for load combinations used for the various parts are given in Table 3.8.3-3. Normal load conditions are shown for mechanical members only. Loads on structural members during normal conditions are negligible and therefore are not shown on Table 3.8.3-3. For normal load conditions, factors of safety for mechanical parts are 5 to 1 on ultimate. For limiting conditions such as SSE and a pipe rupture accident, stresses do not exceed 0.9 yield. For collapsible members during a pipe rupture accident, stresses exceed yield and members are plastically deformed. Plastic deformation of energy absorbing members does not affect the sealing integrity of the doors. 3.8.3.5.11 Seals Between Upper and Lower Compartments Under normal and earthquake conditions, there are no loads on the seals. However, the seals are subject to radiation, as outlined previously, during normal operating conditions. The seal has been tested under accident pressures and temperatures after undergoing heat aging to 40 years equivalent age, and irradiation to 40 years normal operation plus accident integrated doses in order to qualify it for the life of the plant. The seals are not required to maintain their integrity during a fire. It is assumed that a fire and an accident which require sealing will not occur simultaneously since the reactor will be shut down immediately if a fire develops. 3.8.3.5.12 Ice Condenser Table 3.8.3-4 provides a summary of the allowable limits to be used in the design of the ice condenser components. For all cases the stress analysis was performed by considering the load combinations producing the largest possible stress values. Stress Criteria The stress limits for elastic analysis are: 3.8.3-31
WBN
- 1. D + OBE Stress shall be limited to normal AISC, Part I Specification allowables (S). The members and their connections shall be designed to satisfy the requirements of Part I, Sections 1.5, 1.6, 1.7, 1.8, 1.9, 1.10, 1.15, 1.16, 1.17, 1.20, 1.21 and 1.22 of the AISC Specification (stress increase in Sections 1.5 and 1.6 is disallowed for these loads).
Where the requirements of Section 1.20 are not met, differential thermal expansion stresses shall be evaluated and the maximum range of the sum of mechanical and thermal induced stresses shall be limited to three times the appropriate allowable stresses provided in Section 1.5 and 1.6 of AISC Specification.
- 2. D + SSE, D + DBA Stresses shall be limited to normal AISC Specification allowables given in Sections 1.5 and 1.6, increased by 33% (1.33S). No evaluation of thermal induced stresses or fatigue is required. In a few areas, where the stresses exceed 1.33S but are below l.5S, specific justification is provided on a case by case basis.
- 3. D + SSE + DBA Stresses shall be limited to normal AISC Specification allowable given in Sections 1.5 and 1.6, increased by 65% (1.65S). No evaluation of thermal induced stresses or fatigue is required.
For all cases, direct (membrane) mechanical stresses shall not exceed 0.7Su, where Su is the ultimate tensile strength of the material. The summary of the ice condenser allowable limits is given in Table 3.8.3-4. 3.8.3.6 Materials, Quality Control and Special Construction Techniques General Refer to Section 3.8.1.6. 3.8.3.6.1 Materials Refer to Section 3.8.1.6.1 with the following additions. Concrete Aggregates for radiation-shielding concrete which was used in limited locations conformed to ASTM C 637-73. 3.8.3-32
WBN The specified strengths of concrete used for interior concrete structures were 3000 psi, 4000 psi, 5000 psi, and 8000 psi. Reinforcing Steel Prestress steel which was used in the reactor cavity walls conformed to ASTM A 421-65. Personnel Access Doors in Crane Wall ASTM standards were used for all material specifications and certified mill test reports were provided by the contractor for materials used for all load carrying members. Seals Between Upper and Lower Compartments The seals consist of long strips of flexible elastomer coated fabric with both edges hemmed to form pockets into which metal clamp bars are inserted. The coated fabric is two ply dacron coated on both sides with an elastomer (ethylene-propylene-dienepolymer). The elastomer is compound E603 or E603-A by the Presray Company. Escape Hatches in Elevation 756.63 Floor ASTM standards were used for all material specifications and certified mill test reports were provided by the contractor for materials used for all load carrying members. 3.8.3.6.2 Quality Control - Historical Information Concrete The quality control requirements were essentially the same as in Section 3.8.1.6.2. Some concrete did not meet specification requirements. This was evaluated and documented in Reference [2]. Results have been documented in affected calculation packages and drawings. Personnel Access Doors in Crane Wall, Escape Hatches in Elevation 756.63 Floor Design by TVA and erection by TVA were in accordance with TVA's quality assurance program. Design and fabrication by the contractor were in accordance with the contractor's quality assurance program which was reviewed and approved by TVA's design engineers. The contractor's quality assurance program covers the criteria in Appendix B of 10 CFR 50. Fabrication procedures such as welding and nondestructive testing were included in Appendices to the contractor's quality assurance program. ASTM standards were used for the material specifications and certified mill test reports were provided by the contractor for materials used for the load carrying members. Seals Between Upper and Lower Compartments The flexible elastomer coated fabric used for seals was certified by a qualified rubber technologist as being adequate for the normal and accident conditions. In addition, certified mill test reports were provided by the contractor for materials used for the load carrying members. The seal has been tested by the original seal supplier under contract with TVA. The test was designed to evaluate seal specimens under simulated accident temperature and pressure conditions in a configuration emulating actual plant as-constructed installation. The test specimens, which were fabricated from seal material removed from the Unit 1 containment, 3.8.3-33
WBN were heat aged to 40 years equivalent age, and irradiated to 40 years normal operation plus accident integrated doses prior to testing. This testing process represented the material properties that would exist following a design basis accident at the end of a 40 year plant life. 3.8.3.6.3 Construction Technique - Historical Information No unusual construction procedures were employed in the construction of the interior structures. 3.8.3.6.4 Ice Condenser Structural steels for ice condenser components are selected from the various steels listed in the AISC Specification or Code. When materials such as steel sheets, stainless steel or nonferrous metals are required and are not obtainable in the AISC Code, these materials are chosen from ASTM Specifications. Proprietary materials such as insulating materials, gaskets and adhesives are listed with the manufacturers' name on the component drawings. Material certifications for chemical analysis and tensile properties were required with testing procedure and acceptance standards meeting the AISC or ASTM requirements. Because the concept of non-ductile fracture of ferritic steel is not a part of the AISC Code, and Westinghouse recognizes its importance in certain ice condenser components where heavy plates and structurals are used, such as the lower support structure, Charpy V-notch (CVN) energy absorption requirements are stipulated as shown in Table 3.8.3-5. These criteria apply to the design of the following ice condenser components:
- 1. Ice basket and coupling.
- 2. Lattice frame and columns including attachments and bolts.
- 3. Structural steal supporting structures comprising the lower support structure, door frames and bolts.
- 4. Wall panels and cooling duct support studs attached to the crane wall and walls.
- 5. The supports of auxiliary components which are located within the ice condenser cavity but which have no safety function.
The various candidate materials, i.e., steel sheets, structural shapes, plates and bolting used in the ice condenser system were selected on the following bases: 3.8.3-34
WBN
- 1. Provide satisfactory service performance under design loading and environment and pressure or construction performance.
- 2. Assure adequate fracture toughness characteristics at ice condenser design conditions.
- 3. Be readily fabricated, welded, and erected.
- 4. Be readily coated for corrosion resistance when required.
The candidate materials are of high quality and were made by steel-making practices to be specified by Westinghouse. Principal candidate materials meeting the above bases are listed below. Other materials for specific applications are selected on a case-by-case basis. Sheets Carbon steel sheets are commercial quality (CQ), drawing quality (DQ), or drawing quality-special killed (DQ-SK). The selection of the quality depends upon the part being formed. When higher strength, structural quality sheets are required, ASTM specification A607 is used. AISI Type 409 modified stainless steel is a potential alternate sheet material for the ice baskets. The ice baskets were made from perforated sheet material. The wall duct panels were made from sheet material and the cradle supports from structural sections and plates. Structural Sections, Plates and Bar Flats Structural sections, plates and bar flats are generally high-strength, low alloy steel selected for suitable strength, toughness, formability and weldability. The high-strength low-alloy steels are A441, A588, A572 or A633. These steels are readily oxygen cut and possess good weldability. Bolting High-strength alloy steel Type A320 L7 bolting for low temperature service is used for the lower support structure. Stocked bolting made from A325, A449 and ASTM A354, Grade BD (SAE J429, Grade 8) materials are used for other parts. The above bolts met CVN 20 ft-lb at -20°F, for sizes greater than 1 inch in diameter. Nonmetallic materials such as gaskets, insulation, adhesives and spacers are selected for specific uses. Freedom from detrimental radiation effects is required. All structural welding was in accordance with the AWS Structural Code for Welding, D1.1 (AWS Code). The AWS Code is an overall welding system for the design of welded connections, technique, workmanship, qualification and inspection for buildings, bridges, and tubular structures. (See Section 3.8.3.2, Item 5). 3.8.3-35
WBN The quality of welds for the ice condenser system is based on Paragraph 9.25 of the AWS Code. (See Section 3.8.3.2, Item 5). Resistance welding was in accordance with AWS, Recommended Practices for Resistance Welding, C1.1. Magnetic particle examination was performed on at least 5% of the welds in each critical member of the lower support structure. Magnetic particle or liquid penetrant examinations, where applicable, were performed on 5% of the welds in each critical member of the balance of the ice condenser structure. The welds selected for examination were designated in the Design Specifications. The nondestructive examination methods and acceptance standards are given in Section 6 and Paragraph 9.25, Quality of Welds, of the AWS Code. (See Section 3.8.3.2, Item 5). 3.8.3.7 Testing and Inservice Surveillance Requirements Testing of the interior concrete structures was not planned. A completely independent design has been prepared for divider barrier features in order to ensure that during a LOCA the escaping steam will not bypass the ice condenser. Personnel Access Doors in Crane Wall Periodic visual inspections of the doors are to be made. Parts inspected during the visual inspection are to include all bolted connections, structural members for paint deterioration, latches, hinges, and elastomer seals. The seals are to be inspected for cracks, blemishes, or any other indications of deterioration of the elastomer and for proper seating at the sealing surfaces. Seals Between Upper and Lower Compartments On periodic unit shutdowns, visual inspections of the seals are to be made. Parts inspected are to include all bolted connections, clamp bars, metal to fabric joints, and the elastomer-coated fabric. The seals are to be replaced if they show any evidence of deterioration. Escape Hatches in Elevation 756.63 Floor Periodic visual inspections of the hatch covers are to be made. Parts inspected during the visual inspection are to include all bolted connections, structural members for paint deterioration, latching mechanisms, hinges, limit switches, and elastomer seals. The escape latch seals are to be carefully inspected for cracks, blemishes, or any other indications of deterioration of the elastomer and for properly seating at the sealing surfaces. 3.8.3-36
WBN 3.8.3.8 Environmental Effects The atmosphere in the ice bed environment is at 10°F and the absolute humidity is very low. Therefore, corrosion of uncoated carbon steel is negligible. To ensure that corrosion is minimized while the components of the ice condenser are in storage at the site or in operation in the containment, components are galvanized, painted, protective coatings installed, or placed in a protective container. Galvanizing is in accordance with ASTM A123 or A386. Materials such as stainless steels with low corrosion rates shall be used without protective coatings. Corrosion has been considered in the detailed design of the ice condenser components, and it has been determined that the performance characteristics of the ice condenser materials of construction are not impaired by long-term exposure to the ice condenser environment. Since metal corrosion rates are directly proportional to temperature and humidity, corrosion of ice condenser components at operating temperatures has been assumed to be almost nonexistent. Data available in the open literature does not reflect the exact temperature range and chemistry conditions that are expected to exist in the ice condenser, but does indicate that corrosion rates decreased with decreasing temperatures for the materials and conditions being considered. Although the data in the literature indicated that corrosion of components is not expected, Westinghouse has chosen to employ several preventive measures in the construction of the ice condenser system. To inhibit corrosion, galvanizing is used on the ice baskets. Westinghouse has performed tests which show that galvanized material would not be expected to fail due to corrosion during a 40-year exposure to a 5-15°F ice condenser refrigerated air environment. Other structural members are galvanized, protected by corrosion resistant paints that meet the requirements of ANSI 101.2-1972 (Protective Coatings [Paints] for Light Water Nuclear Reactor Containment Facilities) as a minimum, or were constructed of stainless steel, or self-passivating steel. Heavy plate and structural fabrications may be installed in the blasted and/or bare condition. REFERENCES
- 1. TVA Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A.
- 2. Concrete Quality Evaluation, CEB-86-19-C.
3.8.3-37
WBN TABLE 3.8.3-1 LOADING COMBINATIONS, LOAD FACTORS AND ALLOWABLE STRESSES FOR INTERIOR CONCRETE STRUCTURE LOADING LOAD ALLOWABLE STRESSES COMBINATION (1) (4) D L T0 Ta Pa E E' R0 Ra Yj Yr Ym fs fc Vc (2) (3)
- 1. NORMAL 1.0 1.0 1.0 1.0 1.0 0.5fy 0.45fc' .5(3.5 fc')
- 2. EQUIVALENT TEST 1.0 1.0 1.0 0.67 fy 0.45 fc' .5(3.5 fc')
- 3. EXTREME 1.0 1.0 1.0 1.0 1.0 0.9 fy 0.75 fc' 3.5 fc' ENVIRONMENTAL
- 4. ABNORMAL 1.0 1.0 1.0 1.5 1.0 0.9 fy 0.75 fc' 3.5 fc'
- 5. ABNORMAL 1.0 1.0 1.0 1.25 1.0 1.0 1.0 1.0 0.9 fy 0.75 fc' 3.5 fc'
- 6. ABNORMAL/ SEVERE 1.0 1.0 1.0 1.25 1.25 1.0 1.0 1.0 1.0 0.9 fy 0.75 fc' 3.5 fc' ENVIRONMENTAL
- 7. ABNORMAL/ EXTREME 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 0.9 fy 0.75 fc' 3.5 fc' ENVIRONMENTAL (1) INCLUDES ALL TEMPORARY CONSTRUCTION LOADING (2) fy = SPECIFIED YIELD STRENGTH OF REINFORCEMENT (3) fc' = SPECIFIED COMPRESSIVE STRENGTH OF CONCRETE (4) Vc = THIS IS MAXIMUM ALLOWABLE SHEAR STRESS, CARRIED BY THE CONCRETE, WHICH MAY BE REDUCED DEPENDING ON THE SECTION AND TYPE OF LOADING, REF. ACI 359 AS ISSUED FOR TRIAL USE APRIL 1973.
LOAD NOMENCLATURE: D - DEAD LOADS, OR THEIR RELATED INTERNAL MOMENTS AND FORCES L - LIVE LOADS, OR THEIR RELATED INTERNAL MOMENTS AND FORCES T0 - OPERATIONAL TEMPERATURE LOADS Ta - ACCIDENT TEMPERATURE LOADS Pa - ACCIDENT MAXIMUM DIFFERENTIAL PRESSURE E - OPERATIONAL BASIS EARTHQUAKE E' - SAFE SHUTDOWN EARTHQUAKE R0 - PIPE REACTION DURING OPERATING CONDITIONS Ra - PIPE REACTION DUE TO INCREASED TEMPERATURE RESULTING FROM THE DESIGN ACCIDENT Yj - JET IMPINGEMENT DUE TO FLUID DISCHARGE FROM BROKEN PIPE Yr - PIPE REACTION DUE TO FLUID DISCHARGE FROM BROKEN PIPE Ym - MISSILE IMPINGEMENT LOAD
WBN TABLE 3.8.3-2 (1,5)
SUMMARY
OF REPRESENTATIVE MAXIMUM STRESSES FOR THE INTERIOR CONCRETE STRUCTURE (2,3) (4) FLEXURE SHEAR DESIGN FEATURE LOAD
.(6) fc fs ACTL fs' ACTL fcALL fsALL fs' ALL VACTL VALL FOR LOCATION SEE FIGURES 3.8.3-1 THRU 7 COMB ACTL Vc PRESSURIZER COMPT AT CRANE WALL 4 3.5 49.3 9.1 3.75 54 54 148 168 106 ST GEN COMPT, SIDEWALL AT CRANE WALL 6 2.4 49.2 24.0 3.75 54 54 254 290 106 ST GEN COMPT, CENTER WALL AT CRANE WALL 7 2.3 52.6 15.1 3.75 54 54 82 160 79 ICE COND. COMPT, END WALL AT CRANE WALL 4 3.5 39.9 18.4 3.75 54 54 207 275 177 FILL SLAB EL 702.78 AT CRANE WALL 6 3.1 50.9 13.5 3.75 54 54 306 319 155 FLOOR EL 756.63 AT ST GEN COMPT WALL 6 2.9 49.5 22.9 3.75 54 54 489 530 164 CRANE WALL AT EL 702.78 FILL SLAB 6 2.9 51.1 8.7 5.06 54 54 454 462 67 CRANE WALL AT ICE COND COLUMNS 6 3.6 52.2 12.9 6.0 54 54 776 972 99 CRANE WALL AT ST GEN COMPT SIDEWALL 6 2.4 51.0 10.3 5.06 54 54 681 850 164 REACTOR CAVITY WALL-4.25 FEET THICKNESS 6 1.9 46.7 5.6 5.06 54 54 481 485 147 COMPT ABOVE REACTOR-REACTOR CAVITY COLUMNS 4 1.2 40.4 2.4 5.06 54 54 532 681 97 REFUELING CANAL WALL AT CANAL FLOOR SLAB 6 1.1 34.9 2.2 5.06 54 54 351 473 144 REFUELING CANAL FLOOR SLAB 1 0.6 27.4 0.8 3.04 30 30 163 168 60 ICE COND SUPPORT FLOOR-EL 744.5 4 4.2 49.0 13.0 6.0 54 54 569 594 179 CANAL GATE 6 4.2 50.2 19.1 5.06 54 54 341 347 164 CONTROL ROD DRIVE (CRD) MISSILE SHIELD 6 4.0 50.5 16.7 5.06 54 54 295 311 164 REAC COOL PUMP & LOWER COMPT ACCESS HATCHES 6 2.0 46.1 4.4 5.06 54 54 376 408 164 NOTES:
(1) FLEXURAL STRESSES ARE IN KIPS PER SQ IN (KSI) SHEAR STRESSES ARE IN POUNDS PER SQ IN (PSI) (2) fcACTL, fsACTL, fs'ACTL - THE ACTUAL CALCULATED STRESS IN THE CONCRETE, TENSION REINFORCING STEEL AND COMPRESSION REINFORCING STEEL, RESPECTIVELY. (3) fcALL, fsALL, fs'ALL - THE ALLOWABLE STRESS IN THE CONCRETE, TENSION REINFORCING STEEL AND COMPRESSION REINFORCING STEEL, RESPECTIVELY. (4) VACTL - THE ACTUAL CALCULATED SHEAR STRESS IN THE STRUCTURE. VALL - THE TOTAL ALLOWABLE SHEAR STRESS THE SECTION CAN CARRY TO INCLUDE THE ALLOWABLE SHEAR STRESS CARRIED BY THE CONCRETE AS WELL AS THAT PROVIDED BY SHEAR REINFORCING. Vc - THE ALLOWABLE SHEAR STRESS CARRIED BY THE CONCRETE ONLY. (5) THIS TABLE DOES NOT REFLECT THE EVALUATIONS DOCUMENTED IN REPORT CEB 86-19-C. TABULATED STRESSES ARE FROM THE ORIGINAL CALCULATIONS. CHANGES HAVE BEEN DOCUMENTED IN CALCULATION PACKAGES. (6) FOR LOAD COMBINATION DEFINITIONS, REFER TO TABLE 3.8.3-1.
WBN TABLE 3.8.3-3 (SHEET 1 of 4) PERSONNEL ACCESS DOORS IN CRANE WALL LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Normal operating conditions are as follows: Pressure - Negligible Temperature - 30° to 120° F Radiation - 2.0 x 107 rads for 40 year life The effect of pipe rupture accidents on the doors varied with the location and intensity of the accidents. The three types of pipe accidents producing maximum effect on the doors and conditions accompanying these accidents are as follows:
- a. Accidents Without Jets or Missles Hitting the Doors Temperature 327° F for first hour 225° F for next 11 hours Radiation 8.7 x 107 rads total for 12 hours Pressure 12 psig acting from inside crane wall for 12 hours
- b. Accidents With Jet Hitting a Door Temperature 700°F maximum Force and impact As produced by maximum jet Radiation 4.8 x 106 rads per hour (gamma) 2.5 x 107 rads per hour (beta)
Duration of maximum temperature and maximum force from jet is for not more than 10 seconds and then gradually decreases. Pressure and temperature after maximum temperature and force are as outlined in (a) above.
- c. Accidents with Missile and Jet from the Same Source Striking a Door Temperature 700° F maximum Force and impact As produced by jet and missile Radiation x 106 rads per hour (gamma) 2.5 x 107 rads per hour (beta)
Duration of maximum temperature and maximum force from jet is for not more than 10 seconds and then gradually decreases. Pressure and temperature after maximum temperature and force are as outlined in (a) above.
WBN TABLE 3.8.3-3 (SHEET 2 of 4) PERSONNEL ACCESS DOORS IN CRANE WALL LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES (Cont'd) Potential missiles which the doors were designed to withstand are as follows: Temperature element A, without well, boss, and pipe Temperature element B, with well, boss, and pipe Temperature element C, without well Temperature element D, with well Reactor coolant pump temperature element Pressurizer temperature detector Pressurizer heater 2-inch check valve (born injection) 3/4-inch globe valve (sampling system) (flow transmitters) (pressure transmitters) 3/4-inch air-operated valve (head gasket monitoring) 1-inch manually-operated globe valve (excess letdown)
WBN TABLE 3.8.3-3 (SHEET 3 of 4) PERSONNEL ACCESS DOORS IN CRANE WALL LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES (Cont'd) Structural Door and Frame Assembly Allowable stresses (psi)(1) No. Load Combinations Tension Compression(3) Shear I. With door closed or open: 0.6 Fy 0.6 Fy 0.4 Fy Dead load plus OBE II. With door closed or open: 0.9 Fy 0.9 Fy 0.6 Fy Dead load plus SSE III. With door closed: 0.9 Fy 0.9 Fy 0.6Fy Dead load plus SSE plus 12 psig from inside of crane wall IV. With door closed: 0.9Fy 0.9 Fy 0.6 Fy Dead load plus SSE plus Load from maximum jet hitting doors at 615 psi V. With door closed: 0.9 Fy 0.9 Fy 0.6 Fy Dead load plus SSE plus Load from missile with maximum energy 6900 lb/ft hitting door plus jet from that missile source at 295 psi Mechanical Parts Allowable Stresses (psi)(1) No. Load Combination Tension Compression(3) Shear I. With door closed or open: Ult Ult 2 x Ult Dead load plus 5 5 15 Operator force of 75 pounds II. With door closed or open: 0.6Fy 0.6Fy 0.4Fy Dead load plus OBE III. With door closed or open: 0.9Fy 0.9Fy 0.6Fy Dead load plus SSE
WBN TABLE 3.8.3-3 (SHEET 4 of 4) PERSONNEL ACCESS DOORS IN CRANE WALL LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES (Cont'd) Structural Door and Frame Assembly Mechanical Parts Allowable Stresses (psi)(1) No. Load Combinations Tension Compression(3) Shear IV. With door closed: Dead load plus SSE plus Load from maximum jet hitting doors at 615 psi 0.9Fy 0.9Fy 0.6Fy V. With door closed: Dead load plus SSE plus Load from missile with maximum energy (6900 lb/ft) hitting door plus jet from that missile source at 295 psi 0.9Fy 0.9Fy 0.6Fy NOTES: (1) Listed allowable stresses are for non-collapsible members only. Collapsible members are plastically deformed. (2) Earthquake loads act in one horizontal direction only at any given time and act in vertical and horizontal directions simultaneously. (3) The value indicated for the allowable compression stresses is the maximum value permitted when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls. Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or E when Kl > C 2 Fcr = 2 c Kl r r 2
WBN TABLE 3.8.3-4 ICE CONDENSER ALLOWABLE LIMITS(3) Elastic Analysis Load Mechanical Limit Analysis(1) Test (2) Combination Mechanical and Thermal Fatique (Load Factors) (Load Factors) D + OBE S(4) 3S AISC Part I 1.43 1.87 D + DBA 1.33 S N.A. N.A. 1.30 1.43 D + SSE 1.33 S N.A. N.A 1.30 1.43 D + SSE + DBA 1.65 S N.A. N.A 1.18 1.30 NOTES: (1) For mechanical loads only. Mechanical plus thermal expansion, combination and fatique shall satisfy the elastic analysis limits. (2) Membrane (direct) stresses shall be no larger than 0.7 Su (70% of ultimate stress). (3) For particular components that do not meet these limits specific justification shall be provided on a case by case basis. (4) S = Allowable stresses as defined in Sections 1.5 and 1.6 of the AISC Part I Specification.
WBN TABLE 3.8.3-5 SELECTION OF STEELS IN RELATION TO PREVENTION OF NON-DUCTILE FRACTURE OF ICE CONDENSER COMPONENTS(1) Section Thickness Properties 5/8 inch thick and under over 5/8-inch thickness Energy Absorption Level None required i) 20 ft-lb CVN(2) at -20°F for steel over 36,000 psi yield strength ii) 15 ft-lb CVN(2) at -20°F for steel under 36,000 psi yield strength Heat Treatment None required i) Normalizing Steel can be used in the hot ii) Quench and Temper rolled condition Type of Steel i) Rimmed(3) i) Killed ii) Semi-killed(4) ii) Killed-fine grain practice iii) Killed(4,5) iv) Killed - fine grain practice General Notes: (1) Hot rolled, normalized or quenched and tempered steels are used where applicable. (2) Charpy V-notch (CVN) impact testing shall be performed in accordance with the requirements of ASTM A370. (3) Rimmed steel shall be used only for carbon steel sheet products. (4) These type steels shall be applied for components which remain within AISC Code stress limits for all load conditions. (5) Killed steels for above AISC Code stress limits shall be upgraded by heat treatment, e.g., bolting.
WBN TABLE 3.8.3-6 EQUIPMENT ACCESS HATCH
SUMMARY
OF ALLOWABLE STRESSES FOR DESIGN CONDITION I(2),II(2) III(2) Allowable Allowable Bending stress in structural shapes and 21,600 psi 32,400 psi plates (Fy = 36,000 psi) (0.60 Fy) (0.90 Fy) Shear stress in structural shapes and 14,400 psi 21,600 psi plates (Fy = 36,000 psi) (0.40 Fy) (0.60 Fy) Tensile stress in anchor bolts 19,800 psi 31,700 psi (Fy = 36,000 psi) (0.55Fy) (1.6(0.55)Fy) Bearing stress under anchor bolt end 1,250 psi plate (Fc' = 5,000 psi) (0.25 Fc'(1)) Notes: (1) See Table 1002(a), ACI 318-63 Code (2) I = DL + L1 or DL + L2 II = DL + L1 + OBE or DL + L2 + OBE III = DL + L1 + SSE or DL + L2 + SSE L1 = Live load of 14,000 lb (loaded weight of forklift) L2 = Live load of 15 psi pressure from below (LOCA)
WBN TABLE 3.8.3-7 (SHEET 1 of 2) ESCAPE HATCH - DIVIDER BARRIER FLOOR LOAD COMBINATIONS - ALLOWABLE STRESSES Structural Parts - (Fy - 36,000 psi) Allowable Stress (psi) No. Load Combinations Tension Compression(2) Shear Hatch Closed I. Dead load 18,000 18,000 12,000 Live load at 100 lb/ft2 (0.5 Fy) (0.5 Fy) (0.33Fy) Load from latching device II. Dead load 25,900 25,900 17,300 Live load of 15 psi from below (0.72 Fy) (0.72 Fy) (0.48 Fy) Load from latching device SSE(1) Hatch Open III. Dead load OBE(1) 22,000 22,000 14,400 (0.6Fy) (0.6Fy) (0.4Fy) IV. Dead load 25,900 25,900 17,300 SSE(1) (0.72 Fy) (0.72 Fy) (0.48 Fy) Mechanical Parts(3) (Excluding Springs) Allowable Stress (psi) No. Load Combinations Tension Compression(2) Shear Hatch Closed I. Dead load Ultimate Ultimate 2 x Ultimate Live load at 100 lb/ft2 5 5 3 5 Load from latching device Hatch Open II. Dead load Live load of 15 psi from below 0.72 yield 0.72 yield 2 x 0.72 yield Load from latching device SSE 3 Hatch Open III. Dead Load OBE 0.6Fy 0.6Fy 0.4Fy IV Dead Load SSE 0.9Fy 0.9Fy 0.6Fy
WBN Table 3.8.3-7 (SHEET 2 of 2) ESCAPE HATCH - DIVIDER BARRIER FLOOR LOAD COMBINATIONS - ALLOWABLE STRESSES NOTES: (1) Acts in one horizontal direction only at any given time and acts in vertical and horizontal directions simultaneously. (2) The value given for allowable compression stress is the maximum value permitted, when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls. Kl 2 r when Kl < Fcr FY 1 -
= Cc 1 2 Cc 2 r or E
2 Kl Fcr = when > Cc 2 2 Kl r r (3) Pins and shafts, bolts and nuts, bushings, and seals.
WBN TABLE 3.8.3-8 AIR RETURN DUCT PENETRATION
SUMMARY
OF STRESSES FOR CONTROLLING DESIGN CONDITION DB LOCA - DL +/- SSE Calculated Allowable Bending stress in structural 17,900 psi 21,600 psi shapes and plates (0.60 Fy) (Fy = 36,000 psi) Tensile stress in structural 1,890 psi 21,600 psi shapes and plates (0.60 Fy) (Fy = 36,000 psi) Headed concrete anchors (shear) 17,000 psi 27,000 psi (fs = 60,000 psi) (0.45 fs)
NOTE:
- 1. FOR NOTES, REFERENCE DRAWINGS AND COMPANION DRAWINGS, SEE 41 N716-1 Q
r 4 14
~*
a Q a \ NOTE{UNIT 2 ONLY): e MINIMUM ACTUAL STRENGTH ' i ^ OF CONCRETE FOR THESE e O \ g W COLUMNS IS 480- PSI. ,
% W LEW e.
16
=M77~
.Lft
- e e
REFERENCE DRAWINGS: 41 N10080-18 --- CONCRETE POUR DRAWING UFSAR AMENDMENT 1 WATTS BAR FINAL SAFETY ANALYSIS REPORT REACTOR BUILDING UNITS 1 & 2 SCALE: a,_7 -O* CONCRETE INTERIOR STRUCTURE OUTLINE TVA DWG NO. 41N716-5 RD FIGURE 3.8.3-5
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.3-6
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.3-7
SR9. 4/' MOTEA: FOLD TD&WAL/IA4
&YUW 1L ~BY AYELOIAS 70 LLW£R EM9EDMENr AW40AED nAV/T? EL \ IN THIS ARE4.
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awe 1 1 I Ec 7Y19=~ I-If:~ PL /Z FL/tTroR PLAN -EL 714'-
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- 3. F/ELO rOAOOS7TN0 STUD LOCAT/ONAND IAIJ'PL rOREDUCEA/ROAP REACRW /OM/Nf'm 'OETW EDGEOF CLOSURE PLAND MIRROR LNSULA7/ON.
SUPPORT 4,NO OA /?LUIREO FOR "O WLNG NUTS. NBRY27J SF1fL0 MAY NAYS OPT/ NTO USE PAR/ABLE LENGPIS OFfCLOSURE PL AND ADOf"THO STUDS AS NEEDED rD OBT4LNA0000 F/T.
~ NOTE B:
FIELD TO ORAYD CONC.
- REINF. BARS TO OBTAIN SCALE /J"/=0" BEARNO BETWEEN FL WE C.
8 RE/NF BARS. REFER NCE COVER mx, WE.4'-p4
#:5j'TO E REACTOR CAY/TY u ONG.4/N7//. CS MATERIAL, WELDING DEVELOPED ELEVATION OF INSIDE OF REACTOR CAVITY WALL PER TWI 6-290.
EL 724 " AWPECT/DM PORT AZIMUTHS SHOWN ARE REFERENCED TO THE f REACTOR CLOSURE PL RAO// SCALEJ1110" R/ RP R9 NOrE 0+ 20 '
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x ae w PLf W yl E/THERPL /f E AOZZLE EL 7/B_0"] ii iI i I ~o `~ ~\ ~... ~*` 11 t~K TYPICAL CLOSURE PLATE CONNECTION FOR HATCHED AREAS SEE w ~ 1 L7/BLOZ\ MATER/AL AND FABRICATION BY TVA F/CLO l pai14 NTS UNIT 1 AC DRAWING 48W932 hh SEE NOTES i ('BY FIE O W SEE NOTE C~ + "CLOSURE AL AOZZLE EL 7 ' ~~ fAbZZLE ! \ FOR REACTOR Ba`mE cAWrr SUPPORT SEE ;pr... ~ JlLeryr 8070 WATTS BAR
'51"S COMER ABN4b qf-7 +
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~i /L9" /3' ~ _ * ~; SEE 4BN4/0 G
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-K SCALE J'-/Lo-REACTOR BUILDING
' 0Y UNLT2 F EMBEDAIENTS SEE 4SN427 NOTE £:
CUT THIS ARFA OUT. I-N UNIT 2 293"JJJ COVEF BOX ANO WELD 70 PLjIa/AMa A7 - MISCELLANEOUS STEEL G-G ** FILLET NZO PEN TVA 6-290. CUT I2.7° UA0T2 C/~ SCALE*/=0"
~L TAD:BI' OUT LC1L'4TAON BY FIELD. REACTOR CAVITY EMB PARTS SHEET 3
~ F-F L M Fo r l r cA av A AX OF S n* l N H-H $ Hl-Hl COMPANION OWLS; TVA DWG NO. 2-48W932 RO SECTION E-E $EI-El SCAIE 1/=0 S T OS A O T C U Sr 4BW930,48N93/, FIGURE 3.8.3-7F(U2)
SCALE44=/=o" -ALE =ILO" G 48 W533 L L P
n o"a so EArACTom o~ 0' jCE BETWEEN UN/TS f N=3Aj R(ftb/Amwr/ON FART)
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;SPICEa AFTER MP,jYP M3J+200 i EL 722=2" WELD OF AO%LDVTAL AC MS:fB[f i '
!EC 721=d" A07F B: f REAI FIELD 1D " CUT 2%0R0UT HOLES AS CAWrY GRO//T AFTER .T / B=6'R R£OD. CUT-OUTS TO BE PLACEMENT ~ At /f REPLACED &MY FULL AL PLf+S ~*/f PENETRATION W47D5 AFTER 6MlIr PLACEMENT.
KEY PLAN FIELDTO CUT 2"t SCALE /'*/00' BFF{"NMC ANCR(BY ) AND WELD TO N.TJx200 D T s N SECTION A-A BEAM (32 .4#CRS) DETA/L E (ova NAM) SCALE J'/'O' p70' NOTES: I. THE REACTOR BI&DIARS /S A CAYrSORY Z STRUCTURE.
- 2. A FOR ALLsr,MATYR/AL BY TPA FIELD r0 BE DUAL /TY LFYEL jr PfR 4 PLACESI gPBNP NN UC7/ON SPECIE/CAT/ON #J$-BB/
J. ALL AMTERIAL AND FABR/CATA9N BY rYA F/ELO p'C6PTAs Aorm.
- 4. ALL AMTERIAL SHALL BE FABRICATED AND ERECTED IN ACCORDANCE W/TN THE A/SC CODE.
S. ALL WELDING TO BE /N ACCORDANCE W/7N GENERAL CONSrROCj/ON SPECIFICATION 6 -P9C.
- 6. MCIO/N6 00CUMENTAT/AY AM /NSPECPON TOM /N ACCIRD4MCf N/T# CONSTRUCTION afECEC/F/CAT/0N N36 -BB/.
w 7. ALL CARBON STEEL MELD/N6 ELECTRODES TO BE AWS LOfo A 5./ E70 SERIES.
'fairr B. ALL MATERIAL 70 BE ASTM 436.EXCEPT AS NOTED.
IS #Orr 9 FOR SECT~VS NOT SNOMW SEE 480932.
/0 ALL SERCES MW FILL NNETOATA~W WELDS.
// BACK-UP BARS & FIELD SPLICES EXCEPT AS NOTED NOT SHOWN.
/2. MISMATCH OF YERTMAL PL'S AT NOR/Z. FIELD SPOKE EL 7/B=0' TO BE 4L"ANX.
/J. Rg7C//Rfww FABR/CATION AND oK'LML~NFATiAv DF NOTER/AL BY aWrR9C7W SNAIL CWV,9M M 7W 7W 4W /VAN.
QfTA/L B - MK/B MKZ pEA REt70 EA /MI/T /4.COAK. ANCRS. TO BE ASTAI A/OB AND REOL'/aE OAKY TINT N 7J THEY CONFORM 70 ME' APPROPRIATE ASTM YPSUFACAT/CW. C-C SCALE J!/=0' PLAN EL 724=5j` T A S W LNV/T P ORO NANO DET. C SCALE=/=0' FOR HATCHED AREAS SEE UNIT 1 AC DRAWING 48W933 ___1.__" p PLl(rrP) N p PL I~(71Pf i MP - A~
/ ---INSIDE LIAWR PL NOT SNONW F04 CLARITY ti EREC71M ABO[l 7M3 ELEPATIM Atli TO BE COMPLETED Offli 9+ ~6' I ~6 I g ~~ ~6 VESSEL /S /N MACE ok t S Mfj / EL 7AB'O~ EAMME2fIELO SPLM I
MP C t: TYP IMP C 70P OF .wPPoRT o-
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GAP Dw r0 FRAME DOGGING SHAFT, /j'D MIN, 6 DOOR 717LERAIVCES - ASTL4 A322, GRAOE4/40 N-
~ O.
360'1 O' ODOR .22 FOR U / EMBEDDEO FRAME FLOOR f[ 702.76 a
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(44N27/-4 6'O.C. ; ; e FLOOR EC 745.00
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- I'-O' A/N GAP 70 ~* OR LESS f DOOR 4102/ FOR UNIT 2 22\
SECT/ON A-A FLOOR EL 702.78 SCALE A2=/=0' WEDGES F7 E1 \/
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--- DOOR STRUCTURE FLOOR EL 7/6.00
-'--. - (441V27/-2) r070Afr FLUSH AGANST .SK/N R ---
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f- F
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C\j EACH PROM G CATCH i SFAL BAR k'i --' ~- TO Be PROVIDED WITH CArCH OR
=NOTCH COVER AT Top As READ ~~_ -SEAL DEVICE FORHOLDING LATCH E
TO wss OUSSETaNIT 2 C7NLY) ` /N THE FU((Y OPEN AA/O FOR REMWABLE COVER ASSEMWr SEE 44N27/-3 L.. D
-SKm /2 /~ I ------
4_ _ _ _ FUUY CLOSED POSITIONS PLAN 8-,6 D -D DETA/L B Cwee BAR /} Ad FOR DOORS SCALE 6',/'-0' SCAlf' 6'*/'-0' 22/'00' 299°00' FLANGE BAP STM ABT E EXCEPT NO / 24-299AO' AS SHOWN d NOTED 0, G-6 LYvKE Fw HOLDING ZTCH NO SHOWN SCALE 3'*/'-O" GENERAL NOTES:
- 1. FOUR DOOR AND EMBEDDED FRAME ASSEMBLIES REQUIRED.
- 2. EACH DOOR CPERA13LE BY ONE MAN FROM EITHER SIDE.
- 3. FILLET WELD SIZE 3/8' UNLESS NOTED OTEHRWISE.
- 4. BREAK ALL SHARP CORNERS THAT ARE EXPOSED AFTER FINAL ASSEMBLY.
L B 5. QUALITY ASSURANCE INFORMATION ON 44N271-2 AND TVA SPEC 2284. E h4N6E P/N d BAT.K Of CHANNEL ~ SYAI ABr E AS SHOWN d H-H SCALE 3/'-O' MOTE A/
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/ H inn H FOR H/NGEL'EM&S NOTE A SEE 44N27I-2 FOR DETAILS Or FLOOR EC FOR MW AV WASHER 3 -7C SAW THRUST 2 - 7L7L WAS/dER, BUSH/N6 A BUSHING HOUSING, 4 - 7.. SEE 44N27/-3 --~.
0 412, WATTS BAR BASE R( SEE 44NZ 7/-4 _ FINAL SAFETY ANALYSIS REPORT E -E DETAIL C SCALE 6'*r-O' 4F 'D A S W 4 tu
- W OPP HANZY SCALE 3-110" REACTOR BUILDING COMRW/ON DRAWINGS:
UNIT 2 44N27I-2 , 27/-3 B 27/-4 PERSONNEL ACCESS DOORS FOR HATCHED AREAS SEE THRU CRANE WALL UNIT 1 AC DRAWING 44N271-1 SCALE 1-1/2' - 1'-0' DOORS ARRANGEMENT EXCEPT AS NOTED TVA DWG NO. 2-44N271-1 RO FIGURE 3.8.3-8(U2)
La Z Q NOTCN q"D BENT CJ 4 T!/BE 6 4T q"NOTCH _.............. BAR 3 NOT_.f !N 77,IBES d COVE.9 k' FLANGf~ 5 1 - 36 O T53x3x.250 '-~ q FOR SUPPORT PLATE J I 88-x'7-<TYP IL92
..........................__.......... ___..._~..'!'3 ,-J............. ...2'..-__
73 a "9 "'.250- r DETAIL A Z cOIrR d e G TF B'= 3°3is" C TO C STRUT TL/B£ I u 13 SP,9C£S 31 D HOLES-TNRU "MIN /NSIDf R SECTION A*A i
`-ColTP. Rte }" I TYP EIE'ND 2s2 / 410 ASTM .4922 6RADE 4140 fM,IN ULT TENS/1f-&0 XS!/
I-36 ,!TYP-EYERr THIRD T!/BF f
/'- 9jg ................. .........
1" 3.
. _......_.._3:.Z........._...............................-_ HINGE FIN QE-TAIL A MATEAML AS NOTED SCALE 6`-/1O /6 REO D SCALE 6`*1-0" 3z"0 PLAN 125 Srm ALIT t EXCEPT S AS SHOWN B NOTED SHIM ASTM A563 16 RfOO SL'ACC' 6"*I'-O" SIDE OF R f .4S NOTES.-
HINGE 1. GENERAL NOTES ON 44N271-1. 37"D 2. QUALITY ASSURANCE INFORMATION ON 44N271-2 TVA SPEC 2284. I t3. Tu THRUST WASHER SELF LL/BR/CAT/AG &RONZE 16 REDO SCALE 6,-1,0" 9x.250 8"D 7 FN. ,fir 1.509C7' SCALE I. ? =7'-0" 5rM A&r t EXCEPT AS AD7'£D
£XEPT AS FIN- ~-y,,"X 4S' CHAMFER -
SHOWN B NOTED i % ALL EDGES RUSHING SELF IV8RICAr/.VG BRONZE 16 RE0'11 SCALE 5
. Is TJ'P fI x 45 4'D IN CHAM~FER
.D 7 7"D Ns~' F/r~'12S I
191 C WATTS BAR FINAL SAFETY ANALYSIS REPORT D-D TYP SCALE 6"+7'-0" REACTOR BUILDING fJi LOTCH IN BorraM or Two UNIT 1 C.C BUSHING HOUSING COPIER PLATE ASSEMBLr 4 RE0'D AS SHOWN TUBES /6 RMIO ASTM A36 PERSONNEL ACCESS DOORS 4 REO'D OPP NANO SCALE 6'*1'-0` THRU CRANE WALL DOORS MISCELLANEOUS DETAILS TVA DWG NO. 44N271-3 RB COMPANION DRAWAVOS.44N271-1, 271-2 G27/-4 FIGURE 3.8.3-10
0 a
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'[- V. NOTCH I ,F"D PENT TUB£'rw +X NOTCN .TAR TS 9.3, c50 "'~ -. '!VO7Cli !N TUBES & COVER IP FLANGE-T SUPPORT PLATE O
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< ~, DETA!L A Z Z ` < ) ra"'R Fr. a - 3 C T4 C 5:ROT TU8£ 4 'D F/!IES'Ti!RU U SECTION A-A
`- CIWER R 76PW A92', GRADE 4X0
;MIN UL r MA(Slif-BOX1/
1- 36 1 TYP-£ERY THIRD TUBE I' 9 ~a TO { DOOR RING£ PIN DETAIL A MATER/A[ AS N0T2'D 5CAiE 6'*1~O' 6 pro YALE PLAN 3i lr
- N, I YAY ABT!EKCFPT AS S/1Oivf1 B NOTE.'
5HIA4 ASTM A569 16 RIO D
- SCALE 6'* 1' O' 611.1 OF F B A9 TG NOTES:
? HIVSC I 6ENERAL NOTES ON 44101277.1, 2, QUALITY ASSURANCE INFORMATION ON 444271-2 AND TVA SPEC 2264.
P- $ THPU5Tri'A5HER
.SELF 1015R/C4T1N0 BRONZE 16 R£0'O SCALE 6 - !'O'
.3,..r5D Ib D rN2 flT 51'hI ABr sCA1.F ri'= I =O-EXCEPT AS N07ED EXCFGT A
- itYOW'N A' rz zli BUSHING 511Z I UBklCAT1N0 BRONZE-16 RFO'D SCAT£ 6 1' O' F 0 I'll
" 45 WATTS BAR TYP FINAL SAFETY I~ f.1 OTCH /N ANALYSIS REPORT vF TWO COV R~F' A7T5504&r. 8-B C -c TUBES BUSHING HOUSLNIS 1, 4 E S ~7 C .y 16 RCOV ,A57M ,436 CALF J N SCAIE 6'=1'-O' REACTOR BUILDING UNIT 2 PERSONNEL ACCESS DOORS FOR HATCHED AREAS SEE THRU CRANE WALL DOORS UNIT 1 AC DRAINING 44N271-3 MISCELLANEOUS DETAILS TVA DWG NO. 2-44N271-3 RO FIGURE 3.8.3-10(U2)
0 Z 3 AS REQ*O Q K IXISTING UNUSED BOLT HOLES) IN REACTOR EWIPNENT PIT WRB, 0 UNUSED HAND RAIL MOO TING HOLES TI BE L'i PLUGGED, SEAL WELDED, AND GROUND FLUSH L BOLT MATERIAL Z H WITH BASE METAL. THREADED INTO HOLE Q h-0 0 in U 0 0 TYPICAL REPAIR FOR UNUSED BOLT HOLES CURB MATERIAL: SA240 TP304 BOLT: ASTM A103 GR 98 ASJE SECTION X.7304(PER A MATERIAL DERED AS P8 OR WELDING PURPOSES) O A18f 0?24 4M $ 4.177 0 AWR, G71018. Fy (MDD =SO A4 0 Fli O.mU =.taAw t/8 1 d XYPXO'8'RA1E 649fR/RR 4.9.EE7£ ASWAW O6WA G 1-SX B 1X. A57MAK C72 QtAS1MA1Cg(R B. His NOTE: CHAINS SHALL BE E REPLACED WITH SCAFFOLD POSTS NOTE: CLAMPED BETWEEN STD PIPE RLG POST POSTS' CHAINS SHALL 1-1/2° STD HANDRAIL REPLACED WITH 2U 4 SCAFFOLD POSTS 2° SCH SO PIPE II I
,..¢z CLAMPED BETWEEN I I WW POSTS. 7/18°O HOLE FOR 3/4° 3/8°-18 BOLT AND MIN LF NUT AND JAM NUT G
. L 3 4' MIN T
2° MIN T/CURB PL 3,4* MIN I~ THREADED 2 INSERT,TYPE IIA SECTION E-E (ALTERNATE) REMOVABLE NOTES: D. AS AN ALTERNATIVE, THE FIELD MAY REPLACE THE REMOVABLE SAFETY BARRIER POST AND CHAIN BY U F S A R AMENDMENT 1 FABRICATING REMOVABLE HANDRAIL AND POST SLEEVE BY FIELD ASSEMBLIES USING THE TYPICAL DETAILS ON SEE NOTE 1-1/4'6 STD PIPE HANDRAIL 4SES58 SERIES AND SECTION E-E (ALTERNATE). S?RA PIN X f' W IA SLOTTED DRILL AND TAP FOR 1/4° SOCKET HEAD SET SCREW HOLE IN NUT SPRIfH' PIN. CCffi!JE'RCIAL GRADE W. PIN MUST NOT PROTRUDE FRCBf E. ASTM A108, SCH 40 IS AN ACCEPTABLE SUBSTITUTE WATTS B A R AND ROD
%p'p X0'-1' ~,*+,_,,,* B/8°4 HOLEINe3/8.
DRILL a TAP 1/2-13UNC X 5/8° NUT AND HAVE A TIpIT FIT IN ORDER TO STAY IN PLACE. FOR HANDRAIL MATERIAL. FINAL SAFETY F. HANDRAILS, SELF CLOSING CATES. GRATING, KICK DEEP IN CURB BAR 3/4. BY TVA FIELD NOTE A PLATES, LADDERS, AND ASSOCIATED HARDWARE MAY BE PROCURED AND FABRICATED CONFORMING TO SEISMIC ANALYSIS REPORT NOTE. Mr II/I CLASSIFICATION (TVA CATEGORY I (L)). SLEEVE AND SET SCREW ARE OPTIONAL WHEN USING OPTIONAL ASTM A240 TYPE 304
^^ REACTOR BUILDING DETAIL D-O UNIT 2 MISCELLANEOUS STEEL 4 ASTM A276 TYPE 304 REACTOR WELL HANDRAIL & MISSILE SHIELD ANCHOR BOLTS-EL. 756.63 FOR HATCHED AREAS SEE TVA DWG NO. 2-48N940 R3 OPTIONAL D-D UNIT 1 AC DRAWING 48N940 FIGURE 3.8.3-14(U2)
O z 3 Q D: D H W_ 2 Q Z IN"Z J X~V F Q Q U ORATING, HANDRAIL Aim KICK PL SHDWN FOR CLARITY,, m EL 788'-0-1/2* a J*X 14* SEE MOT H-H D CTOe 1fB4" 8zg~ a J*X 7-
~R1 A oDNTOUR END -CT A7-A7 OF 14* WIDEEL ° AS REQ'D NdINIMMIZE 788'-7-1/'
FOR LOCATION, FIT upGAP SEE DET C WITH EMBMEDDED 48N928 STEEL. DET M PLAN VIEW nl lf BNeZg~/
`
SEE DET MM 48W928-1 ~I'~N9291 SEE NOTE 32
&Q PIPE v
~ /KI U T 7CIC PLA'(E NE7 tAIELD71'flCUT FOR)
LADDER ACCESS) r-7KpI/ppCpFPL~ARTTH pE (IXISi UFSAR AMENDMENT 1 I NG p
$ SCR INSTALLATION HATCHED AREAS ARE UNDER WATTS BAR qQ~E UNIT 2 CONFIGURATION FINAL SAFETY 64aN'RfiRG$ 8?j&qN CONTROL SEE 2-48N927-1 ANALYSIS REPORT ALT.SECTION CCCC REACTOR BUILDING UNIT 1 MISCELLANEOUS STEEL FRAMES,
} GRATING AND EMB PARTS, tEL 781'-3*
EL 756.63 - SHEET 1 TVA DWG NO. 48N927 RK J2. THE fLOOR COATING IS TO RE REPLACED, 7HER MW ANY STALL GAPS ALONG THE CONGRE~ STEEL INTERFACE /ILL BE FILLED AT THE TILfE IN ACCDRDAACE
/ITN E NEEar S'ECIFICATION XJA-9J2. 2-fdN927-1,2-48192Q,fdN929 FIGURE 3.8.3-1 5
0"MIN y4"MIN *fe 2-12
)~"MAX "MAX We 2-12 REMOVABLE ACCESS HATCH TO LOWER COMPARTMENT GRATING ~ UF' REMOVABLE GRATING, ATTACH GRATING WITH GRATING FASTENER ASSEMBLY MEL NO. GG-1C 1)¢* REMOVABLE STANDARD GRATING EL 757'-1° EXISTING HANDRAIL (48N940) BY GRATING FASTENERS, LLC OR ENGINEERING 04 KICK a Y4x8 APPROVED EQUAL (FIELD TO CUT KICK a AS S a STD HANDRAIL FL EL 756'-7° EL 757'-1° SEE RAILING DETAILS ftEQ'D FOR LADDER ACCESS) 48E856-4 AND 4SES58-8 Tpp~T~} TO *do *pb; C4 Rm ELEC BDX c ' -f HOLE FOR etx21fx4'-6*t, WITH Q_.o. PL EL 756'-7*
FTELO m DUT aPENO& ix *° %'o WEDGE BOLT °1x214° HD CONC ..~.p: GRATING Pat NN ENDER CGTIROL O 1'-0* MAX C TO C ANCHORS 1'-0° OC BOX PMAx1 C.EA6ASEALL BB1 BB1
- 7.25 m BB1-BB1 As°e
-0*tt* BN v REMOVABLE e 1Y"=1'-~" OVERHANG C4x7 25 GRATING
~EXI NG e _M EL 757'-1° CC1 -CC 1 sue'" 9 2-12 CUES Y C4 lV-1 '-0' ACCESSIBLE PL EL 756'-7°
- day. :day w(x
- Q.o.* :p.o, d rc ~ NEW 1S° FACE OF ~(e CUT 70 PIT DDt OPENING . 90 HOLE FOR dx2x.2 BEVELED SHIM, THK d! ~`
_ WIDTH AS REQ'D x 5° LG 6°tt° 9E°1 WEDGE BOLT
*1'-0° MAX C TO C M AA 1 -AA 1 LL Exisr sTRUCT . J gyti 17¢'=1 '-0" EL 756'-7YA° z°(TVP)fz^
K7~ ~= DD7 W MAKE TANGENT
° C FLOOR EL 758'-7Y,*
EXIST STEEL
~p EXIST. EMBEDDED Asxsv r/coNc. SECTION TT1 -TT1 MAKE L TANGENT O a %4°W HOLE FOR W4x13 a %'s WEDGE BOLT STEEL <"*I -6*t G.C. SECTION V1-Vl ti 1 NP CONTAINMENT
_-- 4 DETAIL A 3. 7'-2* R TO. ,*~7'-0* y, m' 25 / SCALE 2YBN928) CONCRETE OUTLINE R 1.5 // I BELOW GRATING 57'"2* SECTION B-B N.T.S EXISTIN
/ I Xlac P M' 6` CU I
BOUNDARY PERSONNEL ACCESS C R (0` gg(( ABANDONED
.2Y HOLE 0 tES
° rD M~ C1 MK 6 MM1-MM1 C1-C1
'57'-1 ' SCALE 14'-1 '-D' D1 D1 `p, 1 A
- 1. ALL NATERIAL AND FABRICATION 0Y BECITEL FIELD.
- 2. GRATING GATING IS SL-1-104, CAMINO IISG W/ GRATING LOAD CARRYING BARS CUT SWARE.
SCALE 114'1'-0' 2kP3 PART PLAN - EL 746'-0" CLEARANCE TO THE RENDING DABS TO BE SUCH AS TO PERMIT COMPLETE CMDNQINC IISR CLEATING. py 3. TOWN-UP WITH COATIN& SL-1-104* CARMQLINC II50 AT ALL WELDING AND BOLT HALE SURFACES IN
~(6 SCALE' 1'-0' THE GR,ITIM).
51 (UNIT 2 OPP HAND) 4. ALL GRATING TO BE RECTANGULAR TYPE WITH LOAD BARS 114'. Rf~ AND SPACE AT 136 . O.C. 3MIND 3(6 BARS TO DE 114*X If,' AND WELDED CONTINWIS TO EVERY 4th LOAD CARRYING BAR.
- 5. HAND RAILINC TO BE 11PI STD. PIPE FIELD PAINTED.
R. MATERIALS SHALL MNFa N TO THE FOLLOWING: TYP z A. STRUCTURAL SHAPES AND PLATES: ASTM A36 (FW a 38 k.1) OR ASTM A572, GRADE 50. TYP UNO PLCS B.PIPES FOR HANDRAILS: ASTM AN. GRADE 5 SCHEDULE 40. e R11 C.TUBE STEEL: ASTM ABO1, GRADE H. Re D. BOLTS: ASTM A325. SEALANT SEALANT E. NUTS: AMA563, GRADE ON.
/ I \-EXISTING F.9FNNtT RADIUS EL80WS: ASTM A?34. GRADE WPB SCffOa.E 40.
TS3x2 WP 1'-10° NOTCH G.HANDRAIL HOLTS: ASTM A307, AS SPECIFIED ON NOTE 13 OF DWG. 48E986-+. L N1EW M° PLATE {LENGTH AND 1N°(A7) I NEW 71*x7* CURB PLATE 3(6 H.7E* am TAR: I. GRATING: ASTM A7011, GRADE 38 TYPE R. ASTM A7011. HEIGHT AS REWIRED TO ENSURE LEVEL FOOTING AREA) MAX AZ t\LISTING PLATE S1 in EL. 758'-714* 7. REMML FIELD FABRICATED SELF-CLOSINS SAFETY DEVICES ARE DETAILED ON DRAWING 4BES".
- 8. PREFABRICATED SELF-CLOSING SAFETY DEVICES APPROVED BY SITE ENGINEERING INCLUDE F"hE::C"'*
EVERY SECOND LOAD BAR SIDE ONLY PLAN DETAIL Ul - NORTH ,,vv~~ OKs° NEARED SELF-G.OSINT. SAFETY GTE OR INTREPID INDUSTRIES BAR TYPE ATE. oTfffN SELF-G.OSING SAFETY DEVICES APPROVED DY SITE (INENGINEERING OR THE SAFETY ENGINEER MAY BE USED. z CONC. ANCR. O PLAN DETAIL T1 - SOUTH , 2-tz a _ +'-6. O.C. EA 9. COPE TS OR ADD SRIMS/IEVELED SHIMS MERE NEEDED IN ORDER TO INSURE THE GRATING IS LEVEL.
- 10. HANDRAILIND AND GRATING ARE QUALITY RELATED. STRUCTURAL STEEL SHAPES, TUBE STEEL. DOTS
< MIN AND NUTS AM SAFETY RELATED.
(MIRRORED VERTICALLY) c 7fe°W HOLES 1" S~ T
"~~
8 4"
- 11. GRATING FASTENER ASSEMBLY MODEL S1. CT-W WITH SELF TAPPING SCREWS BY CRATING FASTENERS LLC MAY HE USED IN PLACE OF MODEL S). GD-lC, IN AREAS WHERE CRATING IS TO HE ATTACHED TO Yp TUBE _L.
(t EL 783' 'a 12. GL VERTICAL HEIGHT ACCESS LADDER SYSTEM KIT (STAINLESS STEEL) MAY RE USED AS THE OF ~,* SAFETY CLOSING DEVICE. ATE ,a;~ I v iS. CATEGORY II !GETS (DESIGED BY CAT II) DO NDT PROVIDE SUPfW T0, ARE NOT IN THE LOAD IYP) `~z I g *,R _: PA NOTCATEM I STRCTURE, SYSTEM OR D TAC RES TOY LODATING 9AF~ REATED FEATUWIT FEATURES IN Y4" : PROxDDTY OF THESE MEMBERS IS NOT PERMITTED WITHOUT ENGItEER11C APPROVAL. SEE d GLa9ATICIWM6 2_NH9ANDSPECIFICATION NSCr941. 1+. FOR SECTIONS AND DETAILS NOT SCINt SEE 49N928 AND 46N929. T/CONC 1S.pthtERAIL AND RAIL CLIPS FURNISHED BY WESTINGHOUSE. v 1><x5 xz'-414° e EL 756'-7° o 18. ARROW(~ INDICATES DIRECTION ff BEARING CRS.
,y17. ALL FILLET WELDS TO BE X, EXCEPT AS NOTED.
KK1-KK1 GRATING* HANDRAIL AND KICK e SECTION R1-R1 }3L+8NS09MK EL 781'-3° THIS DRAWING SUPERSEDES CONC ANON NTS NOT SHOWN FOR CLARITY PER : ECTION Al-Al DRAWING AD 45NS27 R19 JJ1-JJ1 (NEW) ,*~,'-0* UFSAR AMENDMENT 1 3*-1'-0" HSS3x2 1J~ REMOVABLE ORATING* ATTACH GRATING A WITH GRATING FASTENER ASSEMBLY MODEL DETAIL Kl MIRRORED NO. GG-1C BY GRATING FASTENERS, LLC OR SPIRAL STAIR BASE PLATE EL 756'-7° ENGINEERING APPROVED EQUAL + ~'D1'-0* a.:E VY A7CH EXISTING WATTS B A R (NI CONTOUR END OF 7° 1, HS TOP OF GRATING EL 757'-1° 3f," OVERHANG WIDE a AS REQ'D TO HEIGHT FINAL SAFETY e IV r-EL 786'-TD4* MINIMIZE PIT-UP GAP WITH EMBEDDED STEEL a TO e ANALYSIS REPORT W4 COPE W4 CONTOUR END 14° 3(6
. 754'-1114° WIDE a AS REQ'D TO y.x,+. ,
EXIST. - ----- G6 MINIMIZE FIT-UP GAP Y°x7° e EXISTIN (48N935 _ ` TS 3x2 WITH EMBEDDED STEEL SEE SECT H H (2 48N928) SEE SECT At-A1 REACTOR BUILDING e TO TUBE WELD.,: EXISTING C Z I ,TO ,
~EL 756'-7° (EVE' UNIT 2 aWe EA. SIDE TUBE CTR.
(TYP) e '_ z° J4° d GRATING (48N928) EMBED x7°, FOR TION MISCELLANEOUS STEEL FRAMES, (+t* °' LOCASE L 9 l3x3 GRATING* HANDRAIL AND KICK a NOT SHOWN FOR CLARITY f SEE SECT Al-A1 (AWE) (48N928) FIELD tiA3 OPTION TO GRIND BOTTOM GRATING AND E M B PARTS, HH1-HH1 GRATING, HANDRAIL AND KICK a NOT SHOWN FOR CLARITY FIELD TO DECIDE IF a IS WELDED TO DIAGONAL OR PERPENDICULAR TUBES IN SHOP ~EL 7116'-71¢* NOTE: OF GRATING AS SHOWN TO ALLOW ACCESS DOGR TO OPEN FREELY. E L 756.63 3'=1'-0' JJ1'-JJ1' DETAIL L1 GG1-GG1 DETAIL M1 FOR LOCATION DETAIL N1 FORXISTNG TVA DWG NO. 2-48N927-1 RO FIGURE 3.8.3-15(U2) 3*-1'-0* 3.=1 :_0:: PLAN VIER PLAN VIEW (4 Nt928I ON ANGLE. (MIRRORED VERTICALLY)
Z 3 K O O Z R. FOR POUTING DETAILS. 8fifi NOTfi 8 ON DRAWING 41M36-1. a R2 ~ JR2 a 0 a U FOR HYDROGEN PPE LOCATDN SEE(4TW915b)1~ Rat 2-12 r Y.'Y1D**q nG EL 7Y;' a NOTE At L3x3;a as@t'b'YO.C. APPLY WATER TIWT SEALANT TO LODATIMS WHERE WELD IS NOT AD ER3110. (FWt EXMW'LE A~NNO "HD )P DTTTM OF NEWKATE. 17g SECTION N-N NTS J7 1 NAX 2"(TIP) WS TOS EL 75 R" WS WHEREACCESS (C Y'YY;i 5**LG PER UTS WALL PLATE (2 MXD) 3R6 e 1** 3R6 NEW 1N"CURB PLATE ~FI I ff ADDED IN"PLATE 1*-2**WDExLGAS REOV f TYP T c u. `P y (TYTrP)
//~~ c6n3 )4**sTre NAY F~ LC_. 3)i" NSIFS REIMN EL 756'-012' (TYP)
'Y'IY*"SLOTS FR "B 2"-1-121 3/16 2-12 (TYP) A325 BOLTS (TIP)N C6x13 EXIST 1N"R MET ENBED NN Y"STFFENERNAYRENAN SEE NOTE"N' L3Y3X1/4WTTH RASE PLATE C ONCANCHR SECTION H - H Y*"STEP N.T.S J7 SECTION J1t11 W4 MET.CURB R SECTION J J N.T.S TIP 1** R TO 1** R CUT W4 1 R LOCATE CURB R NS Y* TO FIT N.T.S AS REQ'D WYITH BAR 1/4 1% NOTCH OR FS y NS RFSR 1**R RESPECTTO CURBS AN ptINN SYDOON 1/4* MET.CURB R MR AS NEEDED o cuReR cuReR
~~W' W~
NEWY*" CURB PLATE
- T2Y" LONG SECnONR2a2 NOTE. -- SECTION W1 -W1 (96fi NOITi 'C'j FLL WfIH APPROVED N 2'-0'CURB PLATE TM CURB RAS N.T.S SEC LONG REACTS BLDG SEALANT ORWT
.~, ~LEWD.
PLAN TO SECnR R1R1 REOD TO FO 1" R C~7 SECTION W-W iz %a"=1'-0" pOITT AFTOt Et EL 756-0)i**(SECTDN R2R2) ( !AIR GENERAL NOTE; N- ELT%-7ji*'(SECTIDNRIRI) a
- 1. APPLY WATER TIGHT SEALANT TO LCCATIONS WHERE WELD IS NOT ACCESSIBLE (FOR
- PLATE OVERLAP EXAMPLE AROUND HANDRAIL POSTS) AND ON NMTH BY THE BOTTOM OF NEW PLATE.
NOTCH TOP FLANGE TYP W 9IOE9 aF a UP To 1 *xR* Ys 1 ° R CUT TO FIT TYP FOR ALL G FOR WELD W e KICK
%I-`urn- ACCUMULATORS ECTIONS 3 OPTIONAL WELD DETAIL 3 CURBING IN PLATE ACCUMULATORS 3 d 4 FOR KICK PLATE SPLICES - 1ALT, *,
TYP S SIDES NOTCH TOP FLANGE Ys HESS WELD WATER TIGHT SEALANT. OPTIONAL WELD DETAIL BOTTOM OF NEW PLATE. FOR KICK PLATE CORNERS NOTE 'c'. YKE STUDS TO BE TRINgD BEL NEW A36, X1° PLATE. MATCH ALTERNATE SECTION W1-W1 ALTERNATE SECTION W-W UFSAR AMENDMENT 1 HEIGHT, LENGTH TO SUIT. a 758 1/2- Ifi FLOW. ~ (OTHER INSTALLATION DETAILS (OTHER INSTALLATION DETAILS 1/4*xaa/4* LO AS REWIRED 9R EXISTING yo SHOWN IN SECTION W7-W1) SHOWN IN SECTION W-W) WATTS BAR INDEXING PLATE 7/!*x1° SHORT KICKPLATE -t~ HATTED HOLES FWt EXINTIND NOTE' SNUi TIGHT STUDS. YAMINE INDEXING MARKS JAY NUT TO FASTEN ICK PLATE FINAL SAFETY AS REWIRED BY FIELD. INOE Am PLATE SECTION R-R, R1-R1,R2-R2 tuft ANALYSIS REPORT
, y EITHER LOCATION,
/ 30M Jw" tAl~i* LAvil wTTM I
N.T.S FULL LENGTH PER Zz ON INIDIXINo PLATES). ~(t KICK PLATE WHERE ACCESIBLE FOR HATCHED AREAS SEE we REACTOR BUILDING ASSSSARY UNIT 1 AC DRAWING 48N928 1'J)d TE KICKPLATE MODIFICATION UNIT 2 CURB ACCUMULATOR 3 LADDER MISCELLANEOUS STEEL MMIN MRNTF CRff ~ RAIL PLATE HUTT MULTIPLLET PLATER AjWX RAW OPTIONAL WELD DETAIL FRAMES, GRATING AND F~ TICM FIT FOR KICK PLATE ON CURBS DETAIL G EMB PARTS EL 756.63, SH 2 TVA DWG NO. 2-48N928 R1 FIGURE 3.8.3-16(U2)
,wNfor
-f 34,oFiAl A AAp far J'0 f/af hasded corr ms/Oil cap avew AbtrA .d /"rx* and y-+ pfe'- oc, s/fernafe oN40' (sec )vofe E)
A6AFD~ LIrY/p/y7A,~CfaNva4Ar dap at J, f0c Bar #xf,fA0
.P.talri~~8 ho avt to Atdy Tk/1 .
wshs AArdD(dINC3tiD
.~ FfB andAKLD 1{e(bt r8r /afy C. .
- A!rx~,tAo Nets C:
F&/d to dtifer 1tieNew yvaNet I~r awr PArle where M.ii awmwpiIy Ab Tice of play lassie emhed4bd m Ab camirhi f/ear serface 4r bai/d sp MWffN AA0V0A BOLT) swheeddd f>}rsie /A f/wr sarfaea, as rrfwred,.SbhD
`EXfe d t# A,cow aver ho/b a sA0" " wd tap far tf'I
., fAt Aeaded CRE cap
.m-mv at saeA 04chor ho/t, (see Nete E) KEY PL AN N/ AN TYP 611/DE DETAIL Fa// Scab Note E:
LehytA L thread enyovemewt shall be sOff/cient e wayh to s//om InstalPotion of -gasket materAd and cover P/aces.
/. fl'ELo Add Or, RMaACZ.0 PLAN-PLUG A&8 MWAV /N"A REACTOR COOL ANT POMP Access PLAN-PLUG C NOTE F, E AELO KANDAWA UNIT 2 COVER PLATES @SFA/NN/.WAWAf 3 REND-ACIO A LOWER COMPARTMENT ACCESS AADMaWSSA
/ REOD-piue /*READ ARE @(/AL/rY LEYEL S. A /I/ACLNYAMS MTr N AT AMCA40A BOLTS UNIT 2 PLUS /S OUAL/rY LEVEL Z UNIT 2 aqp NAND lAV/T 2 PLL/es ARf @U9L/rY LEYEL Ir COVER PLATE XMAS 3/18 PLATE, A. ATOCOREAILNT fAARICAr/rN ALA NCYAIEN7AT/IN RF ANTEAW AY COWMAC7AA SMALL CPN/YRAO rP AYE rYA Ad. Ai AN.
&VA, 6"- /=0" A wap/Na DFCO0ENIA7//N AAO /NS/EY7NN RY 70FIELR TIRE AY ALL PIPE TO IN ACCEMANCF RRN N087AOC7ANY SIECIF/CAI/M'NJC-IE/*
ASTM A120 OR 7.rAE MAMW BINLOMN)194 CATE@ORY I STMXriWf. Li AlOS GR 8. ALI A* FAIN ALL AU7EAAE AY 7YA f/ELO re AF fourY LEYEL I MW PLATE TO SE RAMP C"MOMWO SPEC/fIGCT/ON M-OA/ EXCAW At ADTfO. 1 ASTM A36. SEE LAU MATERIAL TO RE ARM Us. EXCEPT AS NOTED. 2-1/2° DRAWING a ALI. WYE To BE am A307. EXCEPF a ww. Sm.40 2-48N923. 0,ALI MATERIAL ADD FAMLCRIOR BY TVA iZELO. EXCEPT PLANS A,R R C. PIPE AEPM SLEEVES TO K AM AND.
&MMKN REOOIRED ISMpOp NER~MIr.
N ATeDIf i .
.iMo1a1S KTO ~A;1TN Ala, 1/" r E S/IpR
-4;E3ISTANEXTCE~EWE Iwcu~
Cu~ KwO tt
. ATALL CDMOSION TO IF AS7M AIRS-
[ I~ YP GRADE W. S/B'aFRAMES i0 K MIItAD 40 TAPPED TO FASTEN OCNER9 ALLIFTIN6 DEVICE NITN FU iNTNE1O CQVMZON KSISFAIF CAP SUM.
./YFCN LDGTia OF WIELD AWES SEE <<N7S6-1.
ALALL NELDZB NT TVA FIEL9 TO K W ACCORDANCE RRN 9EIIERAI IF~ CONETRNCTIOM 3FE90FICATIMI 8.290.
/BALL WELDS M K LARde PENETRANT OR NAM= IANTIME INSPECTED.
.AeELDENG EAECTA=t TO K IS AS.I.F.M SERIES.
PLATE .EISEAL AND BASKET N49RIAL rO RE FININZSKED EY WAVY E0;@NENT i DRILL AND TAP AIIMK r 1-1/2°- 6° TPI AWL A/O COAL AWRS~JLEELRJ A903 ANO NNS/ADNJ r Mr4ff AEMM RAAA/SIAED AV Ali~REtOIBPE CAEY a1W/- Fic4rmV 7mr 7NEY cvAvmw 7D ?w 4fimwmmyr ASTw MWA-MripV. RCP ACCESS OPENING HANDRAIL &ALL MATERIAL SHALL BE FABRICATED AND ERECTED IN I" I fQYAaktO(Ht?Af/AAA DETAIL NTS ACCORDANCE WITH THE AISC CODE. SEE NOTE 24 24.FOR DETAILS OF HANDRAIL POSTS AROUND RCP ACCESS OPENING* SEE 6V &F' /AN/r/ISNJ") 20 Fj=IRED FOR 62126A. FOR ADDITIONAL DETAILS REFER TO DRAWING 2-48N940. FRAME-EL 756=7P' FRAME-EL 756'7j' 28.FOR LIFTING DEVICE FRAh ES WITH DAMAGED 3/8° THREADS, FIELD 4 READ-MN/ /REQD-MAP 9/16 MAY . DRILL, TAP, AND INSTALL 3/8°0x16 THREADS PER INCH. MNELIM CAT I O CLH084X USING DRILL AND TAP PROVIDED IN KIT. USE LOKTITE 242 ON COIL OD, STAKE AT LOCATION AND GRIND COIL FLUSH. Able A.- Am Imiob Awe N Bi prewnf cammfe -- Asr, R//kg o*d a1Ni,Lp p/aaimenf+ a Mote B: /edge cart w dab a .RNA e'7= ~fied /msr ^(a* B#' /;.as m My vd. Redoes depth of ad raves w UFSAR AMENDMENT /raw M iie/d 1r yrmd /j'si"Melia ,@, Arr*was N/'fr wrmAiN Me# doep sew me"- Field to fried order farlate -ro tieiAYaAN Cover 2mo,W WATTS BAR play Midge a srA of#1, as rsfwnaT P'f / cant MeherS** rAfeh to M ride end #'d0ea 2 esoh an OADCO fe sties Cf goarAV FINAL SAFETY bend Dow. owjww ANALYSIS REPORT a.a r~arAve~J FOR HATCHED AREAS SEE aL~ EL 75A=/j'-7 UNIT 1 AC DRAWING 48N923 ANCHOR BO T A99Y: REACTOR BUILDING j` /de xr=4"fhw0obfrof 2-bray C
. hex nvfa; 2~ Ix Oc /pAf sAf UNIT 2 p°/r a/sere,araMr e3fxttra=5',
ew.sadded R sxp(xo=B' rfsneoered cnue MISCELLANEOUS STEEL BO Abgd-MM3 arur dao. SHIELD PLUGS & FRAMES SECT/ON A-A B-B EL. 756.63 Sea/. /fR/w BtrsA. Jx/~D" C-C TVA DWG NO. 2-48N923 R1 L/FT/NG DEY/CE FIGURE 3.8.3-17(U2) S/ell./'O"
Note C: fr IXIf 1RYd)OW. DifRd~v11.1 one°trotlen seal old J Fleld y aat/arind a max of 2'-0* 4 cover h02.. .q'd ia°msd Steel -(
/ I 4- Stool con tolnment vessel i
V....Ia-tit e
,I I
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4 ISO 125 C 4-OWER 100c 90 cuppcp COMPARTME~ OMPARTMENT 75 m 6 12 18 24 30 INCHES F WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT WATTSDIVIDER TYPICAL BAR NUCLEAR FLOOR PLANT FINAL SAFETY LOCA TMPERATURE GRADIE 71 S ANALYSIS REPORT Figure 3.8.3-19 Typical Divider Floor LOCA Temperature Gradients FIGURE 3.8.3-19
WBN 3.8.4 Other Category I Structures The Category I structures other than the primary containment, interior structures, and Shield Building are listed as follows:
- 1. Auxiliary-Control Building and Associated Structures
- a. Control Bay Portion
- b. Auxiliary Building Portion
- c. Additional Equipment Building Portion
- d. Waste Packaging Structure e.. Condensate Demineralizer Waste Evaporator Structure Portion
- 2. Diesel Generator Building
- 3. Category I Water Tanks and Pipe Tunnels
- 4. Class 1E Electrical Systems Structures
- 5. North Steam Valve Room
- 6. Intake Pumping Station and Retaining Walls
- 7. Miscellaneous ERCW Structures
- 8. Additional Diesel Generator Building
- 9. South Steam Valve Room The structures are designed as described in Sections 3.8.4.1 through 3.8.4.7.
Evaluations and modifications of the reinforced concrete structures are optionally done using ultimate strength design methods in accordance with the codes, load definitions and load combinations specified in Appendix 3.8E. Evaluations and modifications of existing steel structures and miscellaneous steel and design of new steel members added after July 1979, are done in accordance with the codes, load definitions and load combinations specified in Appendix 3.8E. 3.8.4-1
WBN 3.8.4.1 Description of the Structures 3.8.4.1.1 Auxiliary-Control Building This building and associate structures are multistory reinforced concrete structures which provide housing for the Engineered Safety Feature Systems, etc., which are necessary for the two reactor units. Certain floors in the control bay, the Condensate Demineralizer Waste Evaporator Structure, and the roof of the fuel handling bay are supported by structural steel framing. Refer to Figures 3.8.4-1 through 3.8.4-9 for the general layout and configuration of the structure. Control Bay Portion Structure The control bay portion is a multistory reinforced concrete structure that is built integrally with the Auxiliary Building portion as shown in Figure 3.8.4-8. The structure is separated from the Turbine Building by a 2-inch expansion joint filled with fiberglass insulation which prevents interaction of the two buildings when subjected to seismic motion. The structure was built in one stage, in advance of the Auxiliary Building, and was used initially as a foundation for the construction gantry crane. Structural steel framing in the control bay consists of four steel framed bays of 25.0 feet by 45.0 feet at Elevation 729.0 and the entire floor at Elevations 741.0 and 755.0, with the exception of the two exterior bays on both ends of the building. At Elevation 729.0 the two exterior bays on both ends of the building are pipe run areas and consequently the floor is 1-1/2-inch deep steel grating on steel beams. Reinforced concrete columns were used not only for structural requirements but also to reduce the beam size required. The floor at Elevation 741.0 is a 1-1/2-inch-deep steel grating on steel beams. The floor acts as both the structural support and the horizontal restraint for the cable tray support systems between Elevation 729.0 and Elevation 755.0. An extensive horizontal bracing system is used at Elevation 741.0 to ensure that the floor maintains maximum horizontal restraint capacity with minimum member sizes. The floor at Elevation 755.0, except in the two exterior bays on both ends of the building, is an 8-inch concrete slab supported on steel beams with intermediate columns on the building longitudinal centerline. The main control boards and instrument racks are located on this floor and the cable trays are supported below it. 3.8.4-2
WBN Control Room Shield Doors Two doors located inside the Main Control Room at floor Elevation 755.0 at doorways C-36 and C-54 provide radiation shielding for personnel inside the Main Control Room during a post-LOCA period. The doors normally remain open and are closed only in the event of a LOCA. The doors are manually operated from inside the control room and require no seals as they do not serve any pressure-confining function. The two doors are identical except opposite hand and operate in opposite hand directions. Each door is a rectangular, structural steel frame with a skinplate on each side, thus forming a hollow box which is filled with lead shot to provide the required shielding. Each door is suspended from above by two monorail-type trolleys operating on a standard structural I beam. The trolley closest to the leading edge of each door is of the geared, hand chain-driven type for opening and closing the door. Manually operated turnbuckle dogging linkages are provided at the top and bottom of each door for firmly securing the doors in either the open or closed positions. Equipment Hatch Covers There are two equipment hatches provided at Elevation 708.0. The covers will normally remain closed at all times during plant operation. Each cover consists of an assembly of three welded structural steel sections held together by rows of steel screws with gaskets provided at each joint in the cover. In the closed position, the covers are secured by steel screws around the periphery. In the event that the hatch covers are removed, supports are provided around the hatch opening for the installation of handrails. These hatches were originally designed to be watertight, however this is no longer required. Auxiliary Building Portion Structure The building is a multistory reinforced concrete structure that provides housing for the Engineered Safety Feature Systems, which are necessary to the two reactor units. The Auxiliary Building structure is attached to the Control Building and located between the Reactor Buildings as shown between column lines q and y, and A-1 and A-15, in Figure 3.8.4-8. In the final constructed state, the Control Building portion will act integrally with the Auxiliary Building portion. The Auxiliary Building is separated from the Reactor Buildings by a 1-inch expansion joint filled with fiberglass insulation that prevents interaction of the buildings when subjected to seismic motion. Seals are provided in the expansion joint to prevent the inleakage of either water or air since the Auxiliary Building, at times, serves as secondary containment. Below grade the seals, which consist of a polyvinyl chloride (PVC) material, are designed to withstand external water pressure, possible detrimental effects of the environment, the anticipated horizontal seismic movement of the buildings, and an assumed differential settlement of 1 inch between the buildings without loss of integrity. 3.8.4-3
WBN The spent fuel pit and fuel transfer canal is housed within the Auxiliary Building. The massive reinforced concrete walls and slab are built integrally with the Auxiliary Building as illustrated by Figures 3.8.4-3 and 3.8.4-5. The spent fuel pit is equipped with two gates. The fuel transfer canal gate (0-GATE-079-0005) is used for dewatering the fuel transfer canal for maintenance. The cask loading pit gate (0-GATE-079-0006) is currently not utilized. The cask loading pit gate will be installed in its stored position only. The fuel transfer canal gate is installed or removed under balanced load. The spent fuel gates are illustrated by Figures 3.8.4-69 through 3.8.4-71. Structural steel framing was used to support the Auxiliary Building roof over the area serviced by the main building crane because of the clear span requirements. This area is approximately 223.0 feet long and 80.0 feet wide. The roof is a reinforced concrete slab constructed on metal roof decking that is supported by steel purlins on welded steel trusses. Railway Access Hatch Covers Six hinged covers shown in Figures 3.8.4-10 and 3.8.4-11 combine to close the railroad access hatch opening in the floor of the Auxiliary Building at floor Elevation 757.0 with the six covers in the raised position, a clear opening of approximately 16-feet, 6-inches by 68-feet, 3-inches is provided over the railroad tracks. Spent fuel casks, new fuel shipments, and major items of equipment entering or leaving the Auxiliary Building above the Elevation 757.0 floor must go through this hatchway. The hatch covers and their embedded frame provide a semi-airtight closure and operate in conjunction with the railroad access door to provide an airlock for the Auxiliary Building against a pressure differential of 1/4-inch of water. An electrical interlock system is provided to interlock the operation of the access hatch covers with the railroad access door. Two limit switches, connected in series to provide redundancy, are provided with each hatch cover and arranged to trip when a hatch cover begins to open. The interlocking of these switches with switches on the door prevents the door from being opened when any hatch cover is open or partially open. In like manner, switches on the door prevent opening of any hatch cover when the door is open or partially open. The hatch covers are required to maintain their integrity and Category I function only when closed. When closed, there is no load on the operating machinery and it has no function to perform. Therefore, the operating machinery is not considered as Category I. Railroad Access Door The railroad access door shown in Figures 3.8.4-12 through 3.8.4-15 for the Auxiliary Building provides closure for the access opening in the east wall at the railroad tracks which are at Elevation 729.0. The door and its embedded frame provide a semi-airtight closure and operate in conjunction with the railroad access hatch covers to provide an airlock for the Auxiliary Building previously described. (This door functions as an ABSCE airlock boundary - Door A112). 3.8.4-4
WBN-1 With the door fully opened, the clear opening in the wall is 16.0 feet wide and 20.0 feet high. All new or spent fuel shipments and major equipment entering or leaving the Auxiliary Building by truck or other means passes through this door. The door and door track are constructed of welded steel. The door, rectangular in cross section, is constructed of horizontal and vertical members with diagonal bracing as required for strength and rigidity. The exterior side of the door is covered with a steel skin plate. The embedded frame for the door is constructed of welded steel and is anchored to the concrete. The door seals in the closed position with the side and top seals compressed against sealing surfaces on the embedded frame and the bottom seal compressed against an embedded sill plate. A sloped track guides the door rollers and positions the door so that the top and side seals contact the sealing surfaces only when the door is in or near the closed position. An electric hoist unit opens and closes the door by lifting and lowering it vertically through a slot in the Elevation 757.0 floor. The hoist unit is mounted on the inside wall above the door slot. The door passes through this slot, and extensions of the frame act as guides for the door in the raised position. The area above the floor at Elevation 757.0, occupied by the hoist and the door in its raised position, is enclosed with an airtight structural steel enclosure with gaskets provided on the access covers necessary for servicing the hoist unit and door. The access door and its frame are required to maintain their integrity and Category I function only when closed. When closed there is no load on the hoist unit, and it has no function to perform. Also, the hoist unit has no function to perform relative to the air tightness of the steel enclosure at Elevation 757.0. Therefore, the hoist unit is not considered as Category 1. Manways in RHR Sump Valve Room Two 54-inch-diameter manways, shown in Figure 3.8.4-16, and located at Elevation 698 feet 1 inch in the walls of the residual heat removal (RHR) sump valve room are provided for each reactor unit. The manways provide passageways through the walls of the sump valve room for workmen, tools, and equipment. The doors will normally be closed during plant operation, when reactor containment integrity is required, unless the doors are open to allow normal maintenance or monitoring activities. Although not required for containment integrity, the manways were designed to remain intact when the doors are open during an earthquake to prevent damage to other equipment in the vicinity of the manways. Each manway consists of an embedded steel frame and a welded steel door. The door is secured in the closed position by bolts. The door is provided with slotted hinges to facilitate opening and closing and to allow for compression of the seals when the door is closed. 3.8.4-5
WBN Pressure Confining Personnel Doors This section covers the following pressure confining personnel access control doors located in the Auxiliary-Control Building. Door numbers listed for the doors are the designations used in Figures 3.8.4-17 through 3.8.4-20. The door details for specific heavy equipment type doors are shown in Figures 3.8.4-21 through 3.8.4-23. The door details for the remaining doors are shown on Reference [1].
- 1. The doors for stairs 7 and 8 penthouses at Elevation 772.0, doors A184 and A191.
- 2. The double doors to the personnel and equipment access rooms, Elevation 757.0 (one for each unit) doors A152* and A159*.
- 3. The double doors at the ice condenser equipment room, Elevation 757.0, door A155.
- 4. The double doors to the emergency gas treatment filter room, Elevation 757.0, door A158.
- 5. The doors to the Reactor Building access room at Elevation 757.0 (one for each unit),
doors A156 and A157.
- 6. The doors for stairs 3 and 4 penthouses at Elevation 757.0, doors A154 and A173.
- 7. The double doors to the elevator shaft at Elevation 757.0, door A153.
- 8. The N-line control bay doors at Elevation 755.0 (two double doors with bidirectional pressure requirements), doors C36 and C54, and Elevation 729.0 (two double doors with bidirectional pressure requirements), doors C29 and C34.
- 9. The N-line instrument rooms access door at Elevation 708.0 (single door with bidirectional pressure requirements), door C26.
- 10. The double doors to the heating and ventilating spaces at Elevation 737.0 (one for each unit), doors A123* and A132*.
- 11. The door separating the Additional Equipment Building and the airlock at Elevation 737.0 (one for each unit, bidirectional pressure requirements), doors A183*, A192*, A214*, and A215*.
- 12. The door to the cask decontamination room, Elevation 729.0, door A115.
- 13. The doors in the X-line wall of the cask loading area at Elevation 729.0 (one single door A113*, and one double door, A114*).
- 14. Doors A161*, A162*, A64, A77, A216, A217, A94, A95, A96, A97, A98, A99, A164, A165, A166, and A167.
- 15. The doors to the main steam and feedwater valve rooms at Elevation 729.0 (one for each unit), doors A101* and A105*.
- 16. The double doors at main entrance from Service Building, Elevation 713.0, door A57*.
3.8.4-6
WBN
- 17. Annulus access door, Door A65,* and door to the Reactor Building access room, Door A64 at Elevation 713.0.
- 18. The airlock door to the radiochemical laboratory at Elevation 713.0, door A55*.
- 19. The door in the C-3 line wall leading to the instrument room at Elevation 708.0, door C20 (water tight and pressure confining).
- 20. The exterior double doors at the entrance to the Unit 1 Additional Equipment Building at Elevation 729.0, door A117.
- 21. The Auxiliary Building door that separates stairwell no. 11 from the Unit 1 ventilation and purge air room on Elevation 737.0, door A125*.
- 22. The Auxiliary Building door that separates stairwell no. 10 from the Unit 2 ventilation and purge air room on Elevation 737.0, door A130*.
- 23. The Auxiliary Building door that separates shutdown board room A from the personnel &
equipment access airlock (that leads to the refueling room floor) on Elevation 757.0, door A151.
- 24. The Auxiliary Building door that separates shutdown board room B from the personnel &
equipment access airlock (that leads to the refueling room floor) on Elevation 757.0, door A160.
- 25. The Auxiliary Building doors that separate the mechanical equipment room from the HEPA filter plenum room at Elevation 772.0, doors A212* and A213*.
- 26. The Auxiliary Building door that separates the upper portion of the refueling room from the airlock that leads to the Auxiliary Building roof at Elevation 786.0, door A206*.
- 27. The Auxiliary Building exterior door that separates the Auxiliary Building roof from the airlock that leads to the upper portion of the refueling room at Elevation 786.0, door A207*.
- 28. The Auxiliary Building door that separates the upper portion of the refueling room floor from the airlock that leads to the Auxiliary Building roof at Elevation 814.75, A208*.
- 29. The Auxiliary Building exterior door that separates Auxiliary Building roof from the airlock that leads to the upper portion of the refueling room floor at Elevation 814.75, door A209.*
- 30. The Control Building door that separates stairwell C1 from the corridor on the west side of the main control room at Elevation 755.0, door C37.
- 31. The Control Building doors that separate the main control room from the Auxiliary Building at Elevation 755.0, doors C49 and C50.
- 32. The Control Building door that separates stairwell C2 from the east side corridor at Elevation 755.0, door C-53.
3.8.4-7
WBN
- 33. The Control Building door that separates stairwell C2 from the corridor leading to the Technical Support Center at Elevation 755.0, door C60.
- 34. The exterior door at the entrance to the Condensate Demineralizer Waste Evaporator (CDWE) Building at Elevation 729.0, door DE1.*
- 35. The Auxiliary Building double doors at the entrance to the Service Building at Elevation 713.0, door A56.*
- 36. The Auxiliary Building door that separates the Auxiliary Building from the airlock that leads to door A55 at Elevation 729.0, door A60.*
- 37. The Auxiliary Building double doors that separate the heating and ventilating equipment rooms from the airlock that leads to door A123 at Elevation 737.0, door A122.*
- 38. The Auxiliary Building double doors that separate the heating and ventilating equipment rooms from the airlock that leads to door A132 at Elevation 737.0, door A133.*
- 39. The interior CDWE Building doors that combine with door DE1 to establish the necessary airlock at the exterior entrance to the CDWE Building, doors DE4* and DE5.*
- 40. The waste packaging room door that separates the waste packaging room from the railroad access room, double door A111.*
- These doors function as an ABSCE boundary. For airlocks, See Table 3.8.4-7b.
3.8.4-8
WBN The doors are hinged, manually operated type metal doors, complete with frames and closers. The frames are either welded to plates, bolted to the concrete walls, or welded to embedded plates. Both single and double doors are involved. Double doors consist of an active and inactive leaf, with the active leaf being used for normal traffic. Doors A65, A55, C20 and C26 have a single skin plate with horizontal stiffeners. Door A57 is a double skinned door with horizontal and vertical stiffeners. All other doors are the flush type. All doors except A55, A57, A65, C20, and C26 are secured for tornado, annulus pressure drop or flood by means of a normal latching mechanism. Door A65 is secured by use of hand-operated dogs and doors A55, A57, C20, and C26 are secured by a dogging mechanism which is manually operated by a handwheel. All doors affected by tornadoes are secured during tornado watches and door A65 is secured during flood warnings. During normal operation, the doors provide personnel and equipment access. Doors A55, A56, A57, A60, A101, A105, A111, A112, A113, A114, A122, A123, A125, A130, A132, A133, A183, A192, A206, A207, A208, A209, A214+, and A215 are also components of the building airlocks which serve to maintain a slight negative pressure in the Auxiliary and Reactor Buildings. These doors are equipped with electrical interlocks to assure that one of each pair of interlocked doors is always closed except when under administrative control. Doors A161, A162, DE1, DE4, and DE5 are also components of the building airlocks; however, they are not electrically interlocked. Doors A55, A56, A57, A60, A101, A105, A111, A112, A113, A114, A122, A123, A125, A130, A132, A133, A152, A159, A161, A162, A183, A192, A206, A207, A208, A209, A212, A213, A214, A215, DE1, DE4, and DE5 are components of the Auxiliary Building Secondary Containment Enclosure (ABSCE) boundary. These doors will be subjected to the slight pressure differential (1/2" water gauge) needed to establish the ABSCE. Doors C36, C37, C49, C50, C53, C54, and C60 are components of the Main Control Room Habitability Zone (MCRHZ) boundary. These doors will be subjected to the slight pressure differential (1/8" water gauge) needed to establish the MCRHZ. Waste Packaging Structure The waste packaging area is a one-story reinforced concrete structure supported on crushed stone backfill placed in four-inch layers and compacted to a minimum of 70% relative density and is located on the north end of the Auxiliary Building as shown in Figures 3.8.4-3 and 3.8.4-5. The roof of the structure slopes about 24° and consists of a series of precast beams topped by 4 inches of poured-in-place concrete. The structure is separated from the Auxiliary Building by a 2-inch expansion joint filled with fiberglass insulation which prevents interaction of the two buildings when subjected to seismic motion. 3.8.4-9
WBN Condensate Demineralizer Waste Evaporator Structure The Condensate Demineralizer Waste Evaporator Building portion is a two-story reinforced concrete structure that houses equipment necessary for processing condensate demineralizer wastes and for serving as a backup in processing floor drain wastes. The structure is supported on H-bearing piles and is located on the northeast side of the Auxiliary Building as shown in Figures 3.8.4-2 through 3.8.4-9. An access tunnel to the waste packaging area is separated from that structure by a 2-inch expansion joint filled with fiberglass material which prevents interaction of the buildings if subjected to seismic motion. Additional Equipment Building Portion The Additional Equipment Building portion consists of multi-story reinforced concrete structures, one for each unit, which accommodate accumulators for each unit and for the transfer of ice condenser equipment. The structures are located adjacent to the Reactor Buildings and near the north end of the main Auxiliary Building as shown in Figures 3.8.4-57 through 3.8.4-59. Each building is founded on sound rock and is separated from the Reactor Building by one inch of expansion joint material which prevents interaction of the building when subjected to seismic motion. South Steam Valve Room The South Steam Valve Room is an integral compartment of the Auxiliary Building portion of the Auxiliary-Control Building. The room is shown on Figures 3.8.4-3, 3.8.4-4, 3.8.4-8, and 3.8.4-49 through 3.8.4-49c. This compartment protects the isolation valves of the main steam lines, and other safety-related equipment, from the effects of tornados and earthquakes, as well as providing support for the main steam and feedwater pipes that exit from the Shield Building. The room is designed in accordance with the loads, load combinations, load factors, and allowable stresses given in Table 3.8.4-1. Structural steel framing is used to support the roofing and roof decking of the valve room. The metal roof decking is designed to blow off to relieve pressure in the room. Protection of the safety-related components within the room from horizontal tornado missiles is provided by the exterior walls of the Auxiliary-Control Building which includes one wall of the valve room. The other walls forming the room are interior to the building and are not subject to impact from horizontal tornado missiles. 3.8.4-10
WBN Protection from vertical missiles is provided by the Reactor Building shield wall and by multiple levels of structural steel beams. The adjacent Reactor Building wall restricts the angles of possible missile entry. Since the roof of the steam valve room is more than 30 feet above plant grade, protection is required for the 1-inch diameter rod, missile A5 of Spectrum A (see Table 3.5-7 and Section 3.5.1.4). The multiple levels of structural steel beams partially screen safety-related components by further restricting possible missile entry angles. Small slender missiles such as the 1-inch diameter rod are known to be aerodynamically unstable and, therefore, tumble in flight. It is highly unlikely that a tumbling missile could strike safety-related equipment due to the limited pathways through the multiple levels of steel support structures. Therefore, adequate protection from vertical missiles is provided. 3.8.4.1.2 Diesel Generator Building The building is a two-story rectangular reinforced concrete box-type structure that houses the diesel generators and associated auxiliary equipment. Interior walls of reinforced concrete separate the diesel generators into four compartments. The diesel fuel storage tanks are embedded in the base slab. The structure is supported on crushed stone backfill placed in four-inch layers and compacted to a minimum of 70% relative density. For general layout and configuration of the structure see Figures 3.8.4-24 through 3.8.4-29. Diesel Generator Building Doors and Bulkheads The four doors shown in Figures 3.8.4-30 through 3.8.4-32 at Elevation 742.0 in the north wall of the Diesel Generator Building along with removable bulkheads above the doors provide closures for the 11 feet - 10 inches high by 8 feet - 8 inches wide access openings to the diesel generator units. They provide for passage of large tools and repair parts for the diesel generators. The doors are normally closed and latched. The bulkheads are bolted in position and are removed only for major repair of the diesel generators. The doors and bulkheads are covered on the outside of the Diesel Generator Building by precast concrete bulkheads as shown in Figures 3.8.4-27 and 3.8.4-33. Together they protect the generators from damage by tornadoes, missiles, wind, snow, ice, and rain and form part of the security to prevent entry into the Diesel Generator Building by unauthorized persons. See Section 3.8.4.5.5. Each bulkhead above the door is a structural steel frame 4 feet 1/2 inches high by 9 feet - 5 inches wide. It is covered on both sides with a steel skin plate and provided with a crushable strip on the inner side along the top and sides. Turnbuckles support the bulkheads vertically, and they are held horizontally by bolted clamps at the sides and top. Each door is 7 feet 1/2 inches high and consists of two leaves that are manually operated and hinged at the outer sides to an embedded steel frame. The two leaves bear against steel bars at the outer sides, against an embedded angle at the bottom, against each other at the center, and against a steel angle at the top. The bars are welded to the embedded frame and the angle to the bulkhead above the door. 3.8.4-11
WBN Each door leaf is a structural steel frame covered on both sides with a steel skin plate and provided with a crushable strip around its periphery where it bears against lateral support. Both leaves are provided with latches that are operated from the inside only. The steel doors and bulkheads were provided in the original design of the Diesel Generator Building to protect the diesel generators from missiles B1, B2, and B3 of missile spectrum B in Table 3.5-8. In a review of the tornado protection criteria by the NRC in 1975 a determination was made that the level of protection provided by the doors and bulkheads should be upgraded to resist three additional missiles (B4, B5, and B6). The existing steel doors and bulkheads were found to be inadequate for the additional missiles. Therefore, precast concrete bulkheads were placed in front of the door openings to provide the additional missile protection. The precast concrete bulkheads consist of several individual sections stacked into place and bolted in position to the concrete walls. The precast concrete bulkheads are required to be in place when the diesel generators are operable. The precast concrete bulkheads are 14 inches thick which is adequate to prevent penetration from missiles B4, B5, and B6. The 14-inch thickness is not sufficient to prevent some scabbing. However, the steel doors prevent the scabbed particles from entering the generator compartments. In the event the steel doors are open the scabbed particles will not reach the diesel generators due to the separation of the generators and doors. This protective scheme of preventing penetration but not scabbing is necessary due to the desire to keep the weight of the precast sections low to facilitate removal by field personnel. 3.8.4.1.3 Category I Water Tanks and Pipe Tunnels There is one refueling water storage tank for each unit at Watts Bar Nuclear Plant. (The functional requirements for these tanks are discussed in Chapter 6.) Pipes extending from these tanks to the Auxiliary Building are housed in reinforced concrete tunnels which vary in width and height. Refueling Water Storage Tanks (RWST) As noted in Tables 3.2-1 and 3.2-2a, the RWST foundation is classified as Category I. The RWST is a Seismic Category I structure. A storage basin is provided around the tank to retain sufficient borated water in the event the tank is ruptured by a tornado missile or other initiating event. Details of the storage basin and the technical basis for it are discussed below. The RWST has a minimum required capacity of 370,000 gallons. The tank is a cylindrical vessel whose longitudinal axis is oriented in the vertical direction. 3.8.4-12
WBN The end of the cylinder which forms the base or bottom of the tanks is completely enclosed with a 5/16-inch thick flat plate. The base of the tanks sits on a support structure consisting of a concrete slab 57 feet -0 inches in diameter and 3 feet -6 inches thick to which the tanks are attached at 48 anchor points. The perimeter ring plate is grout supported by the slab foundation. The interior is filled with approximately 6 inches of sand on top of the concrete slab, which supports the entire surface of the tank base or bottom plate. The top of the cylindrical section of the tank is sealed at the sidewall roof intersection using conical-shaped roofs whose apexes coincide with the tank's longitudinal axis. A barrier wall is located around the RWST to protect the bottom three feet of the tank. This provides storage for borated water after a postulated rupture of the tank. A steamline break in one of the lines outside of the Reactor Building in close proximity to the RWST is the most demanding event that could require suction from the RWST and also credibly be associated with a rupture of the RWST. For this event a maximum of 20,000 gallons of borated water was initially analyzed by the NSSS vendor to provide the negative reactivity to keep the reactor subcritical. A lesser amount has been analyzed to maintain departure from nucleate boiling ratio (DNBR). The barrier wall, however, is designed to retain a volume in excess of this amount while supplying adequate net positive suction head to all ESF pumps. Figure 3.8.4-36a shows the distance (greater than 3 feet) between the top of the wall and the suction intake elevation. The tank is equipped with an atmospheric vent located at the peak or cone apex of the roofs. The vent is designed to pass a volume flow rate of air that is at least equal to the maximum withdrawal rate from the tank. Necessary precautions have been taken in the design of the vent to assure birds, animals, and/or other foreign objects, including, rain cannot enter the tanks. Tank physical dimensions and other parameters are given in Section 9.2.7. The foundations are shown in Figures 3.8.4-35 through 3.8.4-36C. Load combinations for the foundations are shown in Table 3.8.4-20. Pipe Tunnels The pipe tunnels housing the piping extending from the primary and refueling water tanks to the Auxiliary Building are concrete box-type structures that vary in width and height. Protection against flooding of the Auxiliary Building in case of a tank or pipe rupture is provided by walls that separate the tanks from the main tunnel. The layout and configuration of the tunnels are shown in Figures 3.8.4-35 and 3.8.4-36. 3.8.4.1.4 Class 1E Electrical System Manholes and Duct Runs The manholes and duct runs shown in Figures 3.8.4-37 through 3.8.4-46 house the electrical cables that must remain in operation when flood levels rise above the plant grade and emergency power is required for safe shutdown of the plant. The manholes and duct runs lie in soil overburden that varies in depth from 30 to 35 feet. 3.8.4-13
WBN The Category 1E manholes are rectangular box-type structures of reinforced concrete built below plant grade with an access shaft projecting above the surrounding soil. Category 1E manholes are equipped with watertight covers and sump pumps. A concrete cover is provided for protection from vertical missiles. The duct runs connecting the manholes are continuous reinforced concrete beams with embedded electrical conduits. Duct runs are buried with a minimum of 18 inches of soil cover above them, except near the intake pumping station where they are exposed. Duct runs enter the manholes through sealed openings in manhole walls. A minimum of 6-1/2 inches of reinforced concrete is provided above the embedded conduits for protection from vertical missiles. 3.8.4.1.5 North Steam Valve Room The structure, shown in Figures 3.8.4-47 through 3.8.4-49 is designed to protect the isolation valves of the main steam lines from the effects of tornadoes and earthquakes, as well as provide support for the valves and main steam pipes and feed water pipes that exit from the Shield Building. The structure consists principally of several high reinforced concrete walls anchored into a 7-foot-thick base slab that rests on a grillage of reinforced concrete foundation walls supported to rock. A 2-inch expansion joint separates the valve room from the Shield Building. Structural steel framing is used to support roof decking of the valve room. The metal roof decking is designed to blow off when the internal pressure at the roof reaches 72 pounds per square foot. Protection of the components from horizontal tornado missiles is provided by the walls of the steam valve room. Protection from vertical tornado missiles is provided by a 24-inch-thick concrete awning which covers approximately one-third of the roof, by the Reactor Building shield wall, and by multiple levels of structural beams. The concrete awning and the adjacent Reactor Building wall, which extends 89 feet above the decking, restrict the angles of possible missile entry. Since the roof of the steam valve room is more than 30 feet above plant grade, protection is required for small, slender missiles, such as the 1-inch-diameter roof missile A5 of Spectrum A (see Table 3.5-7 and Section 3.5.1.4). (See also Figures 3.8.4-49B through 3.8.4-49C). The main steam safety and relief valves are located about 21 and 17 feet, respectively, below the decking. Four 30-inch-wide flange beams and numerous 8-inch steel channels serve to partially screen the safety and relief valves from tornado missiles by further restricting possible entry angles. The largest pathway through the wide flange beams and the channels is approximately 1.5 feet by 2.5 feet (two such pathways) in plan area. Small slender missiles such as the 1-inch diameter rod are known to be aerodynamically unstable in flight and, therefore, tumble during flight. It is highly unlikely that a tumbling missile could follow the pathways discussed above without being deflected. Therefore, the main steam safety and relief valves are adequately protected from vertical tornado missiles. (See also Figures 3.8.4-49B through 3.8.4-49C). 3.8.4-14
WBN The main steam, main feedwater, and feedwater bypass isolation valves are located below the safety and relief valves and are further protected from missile damage by five levels of wide flange beams (33-inch to 8-inch size) provided for pipe break restraint and support functions. There are no practical pathways by which tornado missiles could reach these valves. (See Figures 3.8.4-48B through 3.8.4-49C). 3.8.4.1.6 Intake Pumping Station and Retaining Walls Pumping Station The intake pumping station is a cellular box-type, reinforced concrete, waterfront structure founded on bedrock and partially backfilled on three sides. On the land side, retaining walls hold back the fill to Elevation 710.0. Permanent openings are provided in the reservoir side of the pumping station to allow flooding of any unwatered pump wells when the reservoir level exceeds Elevation 690.0. The essential raw cooling water (ERCW) pumps, fire protection pumps, and screen wash pumps are located on the upper deck at Elevation 741.0 above the maximum possible flood and are covered by a roof. This deck is completely enclosed by a concrete wall 13 feet high. A wall also supports the structural steel grillage system, shown in Figure 3.8.4-68, which provides tornado missile protection to the equipment below. The raw cooling water pumps are located on the deck at Elevation 728.0, which is below the maximum probable flood, but are not required for maintenance of plant safety. The mechanical and electrical equipment are located on the floors at Elevation 722.0 and 711.0, respectively. A permanent pedestal crane is mounted above the upper deck at Elevation 754.0 for handling of equipment. The structural outline is shown in Figures 3.8.4-50 and 3.8.4-51. Traveling Water Screens As shown in Figure 3.8.4-52, the screens are of the single or through flow, automatic cleaning type with a nominal basket width of 4.0 feet. The capacity of each screen, with a head loss of 6 inches for a clean screen and minimum water depth, is approximately 25,000 gallons per minute at a water velocity of 2.0 feet per second. Basket travel speed is about 10 feet per minute. Removal of trash and refuse from the basket screens is by water sprays located in the head frame. The drive motor for each screen is sized to start the screen with water at Elevation 737.5 and a head loss of 2-feet, 6-inches. All drive components are rated for continuous duty and are suitable for outdoor service. Timers provided in the control circuits for the screens function to operate the screens for 18 minutes every 60 hours to prevent "freezing" of the machinery parts from nonuse. This provides assurance that the screens are in an operable condition at all times. 3.8.4-15
WBN The heads of the screens, including drive components, are located above the maximum possible flood level. The screens are designed to operate during any flood, including a maximum possible flood, with water to Elevation 739.2 and 5 0 head loss. The four screens are identical with two screens provided for each of the two supply trains at the intake station. Each of the two screens on each supply train has sufficient capacity to screen the total water required for one train. The capacity of one supply train is sufficient to supply all water required for the ERCW during a LOCA. Starting of the screens by pressure switches on the spray water assures that adequate spray water for removal of trash is available when the screens are started. This greatly reduces the possibility of carrying trash over the screens and into the screened water. Watertight Doors There are two watertight doors provided on Elevation 722.0, identified as W10A and W10B. These doors prevent water from the room containing one train of the ERCW System from entering the room containing the other train. Concrete Retaining Wall The earthfill is hold back by two concrete retaining walls from the pumping station to a point 32 feet from the pumping station. The concrete walls are keyed into rock and are separated from the pumping station by expansion joints. For outline of walls, see Figure 3.8.4-53. Sheet Pile Retaining Walls The sheet pile walls are parallel and extend from each end of the back of the pumping station toward the main plant. These parallel walls contain earthfill to Elevation 710.0 and project above the sloping grade a maximum of 29 feet at the pumping station. For layout of walls, see Figures 3.8.4-54 and 3.8.4-55. 3.8.4.1.7 Miscellaneous Essential Raw Cooling Water (ERCW) Structures Slabs and Beams Supporting ERCW Pipes At the Intake Pumping Station, the ERCW pipes are supported on a reinforced concrete slab. The slab is approximately 8 feet below grade and 50 feet above bedrock. The slab is supported by a bracket on the pumping station wall, bearing piles, and undisturbed earth. Structural separation from the pumping station is provided by 1/2-inch of expansion joint material. The slab is shown in Figure 3.8.4-56. 3.8.4-16
WBN The ERCW pipes at the Diesel Generator Building are encased in concrete beams for support. The pipes are separated from the beams by insulation and the beams are separated from the Diesel Generator Building by expansion joint material. The beams are supported by brackets on the Diesel Generator Building and by Class A backfill. The beams are shown in Figure 3.8.4-56b. Discharge Overflow Structure The discharge overflow structure is a reinforced concrete box-type structure supported on granular fill material placed over basal gravel. The function of the discharge overflow structure is to provide for the normal flow rate discharge of the ERCW System without unacceptable back pressure if the downstream pipes are blocked and to permit flow to the holding pond under normal conditions. The structural outline is shown in Figure 3.8.4-46a. Standpipe Structures The two standpipe structures are mass reinforced concrete structures placed on firm granular material. The structures have backfill on four sides for the first 8 feet of height and extend 17 feet above grade. The function of these structures is to protect the standpipes from tornado-generated missiles. The structures are shown in Figure 3.8.4-56a. Valve Covers These structures consist of reinforced concrete slabs covering the valves in the ERCW pipes. The slabs are located at grade above the pipes and are supported by either the missile protection slab and/or backfill. The slabs have small openings with precast concrete covers above each valve stem. The openings in the missile protecting valve covers provide immediate access to the valves in an emergency. The structures are shown in Figure 3.8.4-56c. 3.8.4.1.8 Additional Diesel Generator Building The Additional Diesel Generator Building is located 349.25 feet north of the centerline of the Reactor Buildings and 54.5 feet west of the centerline of the Unit 1 Reactor Building. It is a two-story rectangular, reinforced concrete, box-type structure which houses the additional diesel generator unit and its auxiliary equipment. The building is 96 feet long by 53 feet wide and is supported entirely on end bearing structural steel H-Piles as shown in Figure 3.8.4-72. The base slab is 12 feet thick with the finished floor at Elevation 742.0. The diesel fuel storage tanks are embedded in the base slab. For general layout and configuration of the building see Figures 3.8.4-73 through 3.8.4-80. 3.8.4-17
WBN Additional Diesel Generator Building Doors and Bulkheads The two large door openings, shown in Figures 3.8.4-74 and 3.8.4-75, in the north and east exterior walls of the building at Elevation 742.0, provide for passage of large tools and repair parts for the additional diesel generator unit and its auxiliary equipment. Removable missile barriers of precast, stackable concrete sections are installed and bolted into position in front of these doorways to protect safety-related equipment from tornado wind and missiles. These missile barriers also form part of the Security System by preventing unauthorized entry into the building through these doors. Due to the presence of the precast concrete missile barriers in front of the doorways, the equipment doors do not need to function as missile barriers and therefore standard double doors are used. The precast concrete missile barriers will be removed only for major repair of the diesel generator. 3.8.4.2 Applicable Codes, Standards, and Specifications Unless otherwise indicated in the UFSAR, the design and construction of the Category I structures other than the primary containment and interior structures are based upon the appropriate sections of the following codes, standards, and specifications. Modifications to these codes, standards, and specifications are made where necessary to meet the specific requirements of the structures. These modifications are noted in Sections 3.8.3.2, 3.8.4.3, 3.8.4.4, and 3.8.4.6. Where date of edition, copyright, or addendum is specified, earlier versions of the listed documents were not used. In some instances, later revisions of the listed documents were used where design safety was not compromised. 3.8.4.2.1 List of Documents
- 1. American Concrete Institute (ACI)
ACI 214-77 Recommended Practice for Evaluation of Strength Results of Concrete ACI 318-63 Building Code Requirements for Reinforced Concrete. (See Section 3.8.4.2.2 for basis for use of this section.) ACI 318-71 Building Code Requirements for Reinforced Concrete ACI 347-68 Recommended Practice for Concrete Formwork ACI 305-72 Recommended Practice for Hot Weather Concreting ACI 211.1-70 Recommended Practice for Selecting Proportions for Normal Weight Concrete 3.8.4-18
WBN ACI 304-73 Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete ACI 349-76 Code Requirements for Nuclear Safety Related Concrete Structures, Appendix C only ACI 531-79 Building Code Requirements for Concrete Masonry Structures
- 2. American Institute of Steel Construction (AISC), "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," adopted February 12, 1969, as amended through June 12, 1974, except welded construction is in accordance with Item 5 below.
- 3. Steel Structures Painting Council, Surface Preparation Specification No. 2, 'Hand Tool Cleaning.'
- 4. American Society for Testing and Materials (ASTM), 1971 Annual Book of ASTM Annual Standards.
- 5. American Welding Society (AWS)
'Structural Welding Code,' AWS D1.1-72, with revisions 1-73 and 2-74 except later editions may be used for prequalified joint details, base materials, and qualification of welding procedures and welders.
Visual inspection of structural welds will meet the minimum requirements of Nuclear Construction Issues Group documents NCIG-01 and NCIG-02 as specified on the design drawings or other engineering design output. See Item 18 below.
'Recommended Practice for Welding Reinforcing Steel, Metal Inserts, and Connections in Reinforced Concrete Connections,' AWS D12.1-61.
- 6. American Gear Manufacturers Association, Standards for Helical and Herringbone Gears.
- 7. Uniform Building Code, International Conference of Building Officials, Los Angeles, 1970 Edition.
- 8. Southern Standard Building Code, 1969 Edition, 1971 Rev.
- 9. 'Nuclear Reactors and Earthquakes,' USAEC Report TID-7024, August 1963.
3.8.4-19
WBN
- 10. American Society of Civil Engineering Transactions, Vol. 126, Part II, Paper No. 3269, 'Wind Forces on Structures,' 1961.
- 11. Code of Federal Regulations, Title 29, Chapter XVII, "Occupational Safety and Health Administration, Dept. of Labor", Part 1910, 'Occupational Safety and Health Standards.'
- 12. Regulatory Guides (RG)
RG 1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures RG 1.13 Fuel Storage Facility Design RG 1.15 Testing of Reinforcing Bars of Category I Concrete Structures RG 1.31 Control of Stainless Steel Welding RG 1.55 Concrete Placement in Category I Structures
- 13. Section deleted in initial UFSAR
- 14. TVA Reports TVA-TR-1, Topical Report, TVA-TR-1 Protection Against Pipe Whip Resulting From Piping Ruptures, 1973.
TVA-TR-78-4 Design of Structures for Missile Impact, 1978. CEB-86-12 Study of Long Term Concrete Strength at Sequoyah and Watts Bar Nuclear Plants CEB-86-19-C Concrete Quality Evaluation
- 15. National Electrical Manufacturers Association, Motor and Generator Standards MG-1, 1970 Edition.
- 16. Structural Engineers Association of California, "Recommended Lateral Force Requirements and Commentary," 1968 Edition.
- 17. National Fire Protection Association (NFPA) 30.
- 18. Nuclear Construction Issues Group (NCIG)
NCIG-01, Revision 2 - Visual Weld Acceptance Criteria (VWAC) for Structural Welding At Nuclear Power Plants NCIG-02, Revision 0 - Sampling Plan for Visual Reinspection of Welds 3.8.4-20
WBN The referenced NCIG documents may be used after June 26, 1985, for weldments that were designed and fabricated to the requirements of AISC/AWS. NCIG-02, Revision 0, was used as the original basis for the Department of Energy (DOE) Weld Evaluation Project (WEP) EG&G Idaho, Incorporated, statistical assessment of TVA performed welding at WBNP. Any further sampling reinspections of structural welds subsequent to issuance of NCIG-02, Revision 2, are performed in accordance with NCIG-02, Revision 2 requirements. The applicability of the NCIG documents is specified in controlled design output documents such as drawings and construction specifications. Inspectors performing visual weld examination to the criteria of NCIG-01 are trained in the subject criteria. 3.8.4.2.2 Basis for Use of the 1963 Edition of ACI 318 The reason for using the 1963 edition of the ACI 318 Code was that much of the Watts Bar Plant was a duplicate of the Sequoyah Plant, for which structures were designed using the 1963 edition. On that basis, design computations for the Sequoyah Plant were the initial design computations for the Watts Bar Plant. In some instances, structures could not be duplicated and new design computations were prepared for these structures with the designs in accordance with the ACI 318-71 Code. Within duplicate structures, where loading changes required investigation of the Sequoyah design for an element of the structure, and the result was a change in member size or reinforcement requirement, the redesign for the member was in accordance with the ACI 318-71 Code. The duplicate structures are the Auxiliary-Control Building and the Diesel Generator Building. The differences between the two code editions for working stress design were examined and the conclusion was that none of these differences significantly affect the safety of the plant. The following are comparisons of the important parts of the code which affect safety: ACI 318-63 ACI 318-71 Flexure W. S. Design Alt. Sect 8.10 Concrete Stress 0.45 fC 0.45 f'C Steel Stress 24,000 psi 24,000 psi 3.8.4-21
WBN Tied Columns Maximum P 0.212 fCAg 0.268 fCAg
+20,400 As +18,900 As Balanced P ~0.15 f'CAg ~0.16 f'CAg Concrete Shear Stress Beam VC 1.1 f C 1 1.1 f C 2 Slab VC 2 f C 3 2 f C 4 Bearing Stress Concrete 0.25 to 0.375 f'C 0.3 to 0.6 f'C One of the apparent major differences between the two codes is the method by which rebar contact splice lengths are calculated. The 1971 code does require longer splice lengths where bars are spaced closer than 6 inches, but it is TVA practice to specify bar spacing of 6 inches or greater for bars which must be lapped by a contact splice. Listed below is a comparison considered to be typical in TVA practice. The results of this comparison show that the use of the 318-63 Code does not significantly affect the safety of the plant.
Contact Splice Lengths f'C = 3,000 psi fy = 60,000 No. 11 bar in tension, top bar Bar spacing, 6 inches to 17 inches 3.8.4-22
WBN 318-63 318-71 W. S. Design Alt. Sect.8.10 As Provided Equals That Required by Computation Less than 1/2 bars spliced at a given section 102 inches 99 inches More than 1/2 bars Not specified spliced at a given TVA practice section 1.2x102=122 inches 129 inches As Provided Equals Twice That Required by Computation Less than 3/4 bars spliced at a given section 102 inches 76 inches More than 3/4 bars spliced at a given section 102 inches 99 inches 3.8.4.3 Loads and Loading Combinations 3.8.4.3.1 Description of Loads See Tables 3.8.4-1 through 3.8.4-13, Tables 3.8.4-15 through 3.8.4-23, and Appendix 3.8E for the loads for other Category I structures. Other Category I structures are subject to the same natural phenomena and basic dead, live, and earth pressure loading as described for the Shield Building in Section 3.8.1.3. In addition to active earth pressure loading described in Section 3.8.1.3, the other Category I structures are designed for at rest and passive earth pressures where applicable. Construction loads differ for the Auxiliary Building because of the multistory effect of shoring from one floor to the next and the construction crane loading on the Control Building portion. 3.8.4-23
WBN The maximum temperature gradient for walls above grade with exterior exposure is the same as the normal operating temperature gradient of the Shield Building. The spent fuel pit and fuel transfer canal require additional temperature considerations. Under accident conditions the water was assumed to reach 212°F in 8 hours with the inside building temperature initially at 60°F. The normal temperature of the water in the fuel pit and canal is 120°F. Hydrostatic pressure loads in the fuel pit and canal vary with water levels in the pit, cask loading area, and canal. The water in the cask loading area is normally maintained at approximately the same level as the water in the spent fuel pool. The canal may be emptied. The wind and tornado loading are described in Section 3.3. Blowout panels are necessary to restrict the maximum internal pressure to 1.25 psi above the Elevation 757.0 floor in the Auxiliary Building as shown between column lines t and y, and A-5 and A-11 in Figure 3.8.4-4 and 3.8.4-5. The load associated with supports for cable trays, piping systems, and other fastenings to interior reinforced masonry walls was restricted to a maximum of 20 psf over one face of the wall (i.e., 10 psf on each face). A 1730 psf surcharge loading was applied to the A-1 and A-15 line walls as a construction loading in the Auxiliary Building. 3.8.4.3.2 Load Combinations and Allowable Stresses See Tables 3.8.4-1 through 3.8.4-13, Tables 3.8.4-15 through 3.8.4-23 and Appendix 3.8E for the loading combinations and allowable stresses for concrete, miscellaneous steel, and structural steel. The normal allowable stresses of ACI 318-63 and ACI 318-71 were used for the basic loading combinations of dead, live, earth pressure, hydrostatic ground water to Elevation 710.0 (or full pool water levels in the spent fuel pit) and effects of normal temperature gradients. For additional loads such as induced moments or shears resulting from operating basis earthquake (OBE), accident pressure loading caused by a LOCA or steam pipe rupture and thermal effects corresponding to the accident condition, a 25% increase in steel stress was allowed with concrete stresses restricted to normal allowables. For construction loading instead of normal live loading or for hydrostatic pressure to Elevation 724.4, a 35% increase in both steel and concrete stresses was allowed. For the combination of the basic loads with safe shutdown earthquake (SSE) effects, or tornado wind loads and associated missiles, or maximum possible flood loads, or impact loadings from jet impingement or jet loading on pipe restraints in conjunction with accident pressures a 67% percent increase in concrete stresses was allowed with steel stresses allowed to reach 0.9 of yield. 3.8.4-24
WBN The maximum lateral forces generated by the SSE are transmitted to the base through shear walls which are designed in accordance with Section 2631 (c) of the "Recommended Lateral Force Requirements and Commentary," of the Seismology Committed, Structural Engineers Association of California, 1968. 3.8.4.4 Design and Analysis Procedures 3.8.4.4.1 Auxiliary-Control Building Control Bay Portion This concrete structure was designed in accordance with the ACI Building Code 318-63 using the elastic working stress theory. The loads, loading combinations, and allowable stresses used are as given in Section 3.8.4.3.2. The control bay was designed as an independent structure. A standard frame analysis was performed on the building in the design of the main structural walls and a separate analysis was performed for each loading combination. The stage that construction of the building's component walls, slabs, and columns would have progressed by the time of the application of a particular loading, was taken into account and reflected accordingly in the model frame. The floor slabs at Elevations 708.0 and 729.0 were designed by ICES STRUDL-II, Volume I program as flat slabs restrained at the exterior structural walls and supported on concrete columns. At Elevation 755.0 the two exterior bays on both ends of the building were designed by ICES STRUDL-II, Volume II program to resist a break in the main steam lines below. The roof slab was designed as a one-way slab spanning between the walls at column lines n and q, as shown in Figure 3.8.4-8. These walls act as shear walls in the event of east-west seismic motion or any other east-west lateral force, with the walls along column lines C1, C3, C11, and C13, as shown in Figure 3.8.4-3 acting as shear walls for north-south lateral forces. The roof slab and floor slabs act as diaphragms. The columns and main structural walls transmit vertical load to the base slab. The floors and walls of the Auxiliary Building are continuous with the control bay north wall. Dowels and shear keys were provided in the wall in order to provide for this structural continuity. 3.8.4-25
WBN Procedures used to design the structural steel framing were based on simple beam and column construction as covered in AISC 'Specification for the Design, Fabrication and Erection of Structural Steel for Buildings,' Part 1, with type 2 framing connections. The beams at Elevation 755.0, between column lines C3 and C5, and between column lines C9 and C11, were designed to function compositely with the concrete slab through the use of headed concrete anchor studs welded through steel decking to the top beam flanges. The beam-to-beam and beam-to-column connections were typical AISC double angle connections as required by the beam reactions, using either rivets or high strength bolts. Between column lines C5 and C9 the beams were not designed to function compositely. For column line references, see Figures 3.8.4-3 and 3.8.4-4. In these areas horizontal bracing is used to resist the horizontal forces for the support of components such as cable trays, conduit, and pipe supports. At Elevation 741.0, there were special connections required that were either bolted or welded in accordance with the codes, standards, and specifications identified in Section 3.8.4.2.1. Transfer of loads into the concrete structure was through bearing plates. Reinforced concrete partition walls are shown in key plan on Figures 3.8.4-60 through 3.8.4-65. These walls were analyzed as free at the top, fixed at the base, and were designed to resist seismic stresses. A minimum steel percentage of 0.1 was provided horizontally for each face. A 2-inch space was left between the top of the walls and the bottom of the slab or beam above, in order to ensure that the walls do not act as structural components of the building frame. Each side of this space was filled with a minimum of 2-inch-wide grout. All reinforced masonry walls are designed in accordance with ACI 531-79 and NUREG-0800, Section 3.8.4, Appendix A. Control Room Shield Doors The doors were designed assuming that the entire dead load is carried by the two vertical members in the door directly under the trolleys with the load from the lead shot acting as a fluid pressure load. The end panels were designed as a fixed beam with uniform load, while the skin plate was designed as a square flat plate stayed at the four corners. The top and bottom members of the door were considered as simple beams. Earthquakes are the only natural environmental condition which applies to the doors. Being inside the control room, the doors are protected from outside elements. 3.8.4-26
WBN For design, the earthquake loads for the various parts consisted of the loads produced by an OBE or an SSE. Accelerations due to a SSE are greater than those due to an OBE by a factor of 2. The doors were designed to maintain their structural integrity based on loading conditions resulting from an OBE or an SSE. Earthquake loads used in design of door and dogs were the greater loads produced by OBE or SSE having peak accelerations at floor Elevation 755.0 in the Control Building. These accelerations were used as static loads for determining component and member sizes. After establishing the component and member sizes, a dynamic analysis, using appropriate response spectra, was made of the door and dogs. This analysis indicated that additional stiffness was required in order to limit the loads on the dogging linkages such that allowable stress would not be exceeded. After adding diagonal stiffeners, another dynamic analysis was made and it was determined that the allowable stresses had not been exceeded. The door assembly was qualified to the Set "B" response spectrum. Equipment Hatch Covers The covers were designed to resist a downward uniform pressure created by a 3-foot head of water caused by a Condenser Circulating Water System (CCWS) rupture in the Turbine Building in addition to the dead load of the structural steel components. The covers were also investigated for a uniform pressure differential of 3.0 psi upward caused by the rapid depressurization during the occurrence of a tornado. These hatches were originally designed to be watertight which no longer is a requirement. Since the original hatches remain in place, the loads to which the hatches were designed remain the same. The covers were designed as a structure supported around its periphery. The structural steel members were designed as simply supported beams with uniformly distributed loads. Loads from the covers are transmitted to embedded frames which are continuously supported by concrete. The embedded structural steel frames are solidly anchored in the concrete by headed steel studs welded to the frames. Auxiliary Building Portion This concrete structure is designed according to the ACI Building Code 318-63 and the stresses are determined by the working stress method for the principal design cases as shown in Section 3.8.4.3.2. Stresses resulting from the static analysis are combined by the method of superposition with stresses resulting from moments, shears, deflections, and accelerations determined by the dynamic earthquake analysis described in Section 3.7.2. The exterior concrete walls above grade Elevation 728+ are designed to resist the tornado-generated missiles as described in Section 3.3. 3.8.4-27
WBN The condition of rapid depressurization during a tornado is provided for in the following manner. The exterior part of the building is designed for an internal positive pressure of 3 psi occurring in 3 seconds with the following exceptions:
- 1. The area above the refueling floor at Elevation 757.0 as illustrated by Figure 3.8.4-3, is designed with blowout panels which open at 0.25 psi. The roof and exterior walls of the spent fuel pool room and cask loading area were evaluated for the effective tornado-generated pressure differential and were found to be within allowable stress limits.
- 2. The area below the Elevation 786.0 roof is vented from openings in the roof as illustrated in Figure 3.8.4-8. The roof and walls housing this area are nevertheless designed for 3 psi. The floor at Elevation 772.0 below this roof is also designed for an uplift of 3 psi recognizing the venting of the area above this floor.
- 3. The heating and ventilating rooms at Elevation 737.0 (see Figure 3.8.4-3) are vented by the air intakes on the exterior walls. This results in the floor, roof, and interior walls of these rooms being designed as exterior member for 3 psi pressure.
The exterior walls below grade Elevation 728+ are designed for earth pressures. The exterior walls on the east and west ends of the Auxiliary Building are designed as cantilevered retaining walls from Elevation 690 to Elevation 711+. These walls are built early before any adjacent walls and slabs to allow the construction field force to backfill and have early access to the area at Elevation 711+. The lateral earth pressure is calculated using Coulomb theory and values are given in Section 3.8.1.3. The exterior walls north of the Shield Buildings with the buttress walls framing into them, as shown In Figure 3.8.4-2, are designed as cantilevered retaining walls from Elevation 690 to Elevation 727+ to allow for earth backfill and placement of the Elevation 729 slab on grade. Horizontal seismic forces are resisted by shear walls with the floor slabs and roof acting as diaphragms. Only those walls parallel to the seismic motion are assumed to resist that motion. The total shear at any level is proportioned among the shear walls in accordance with the method in Reference [3]. For the Safe Shutdown Earthquake, an allowable ultimate shear stress of 5.4f'c is used. This is the value specified in the SEAOC Code in Section 2631 (c) for walls with a height to width ratio less than one, as is the case for this structure. For the operating Basis Earthquake, an allowable value of one-half of the above is used. Main steam and feedwater water pipes penetrate the exterior walls of the valve rooms on the south sides of the Shield Building. These penetrations furnish pipe restraints through flued heads embedded in the walls. The flued heads restrict concrete surface temperatures to a maximum of 150° F. 3.8.4-28
WBN The primary structural support system is designed as a flat slab floor system with concrete columns. Large openings that required separate design are framed with beams. The thickness for many slab sections throughout the building is determined by shielding requirements. The general thickness and live load requirements for different slab areas are shown on Figure 3.8.4-9. The major portions of the building slabs are designed by the ICES STRUDL-II, Volumes I and II. Moments and shears from small frames, beams, and one-way slabs are designed by the moment distribution method. Where slabs act as two-way slabs due to walls or beams below, moments and shears are determined by use of method 2 of Appendix A in ACI Code 318-63. The effects of the relative deflections of the supports and the effects of column shortening were taken into account in the design of all slabs. The minimum percentage of reinforcing in the slabs is 0.15% in the top face and 0.18% in bottom face. The roof slab at Elevation 786.0 is designed for 3 psi uplift pressure as a flat slab using the ICES STRUDL-II, Volume I computer program. The roof at Elevation 801.0 is also designed for 3 psi uplift pressure using the ICES STRUDL-II, Volume II Finite Element Method. In the interior of the building there are many areas around equipment that require shielding which is provided by poured-in-place concrete walls. To permit equipment installation the construction of shielding walls is delayed until the building frame and floor construction is completed and equipment is installed. These shield walls contain minimum steel percentages in the horizontal and vertical directions as specified by the TVA Temperature and Shrinkage Standards and the ACI Code 318-63, Section 2202 (f). These walls were checked for stresses resulting from seismic loading; however, seismic stresses did not control. Reinforced concrete partition walls are shown in plan on Figures 3.8.4-60 through 3.8.4-65. These walls were analyzed as free at the top, fixed at the base, and were designed to resist seismic stresses. Minimum steel percentages provided were the same as those described for the shield walls. A 2-inch space was left between the top of the walls and the bottom of the slab or beam above in order to ensure that the walls do not act as structural components of the building frame. Each side of this space was filled with a minimum of 2-inch-wide grout. 3.8.4-29
WBN-1 The thick concrete walls of the spent fuel pit and transfer canal are required for shielding. They are shown in Figure 3.8.4-3. The walls are supported by a concrete base slab, which is approx-imately 27 feet thick. The walls and base slab are built integrally with the slabs and walls of the Auxiliary Building. A structural wall separates the cask loading area from the spent fuel storage area. The design of the pool walls take into account hydrodynamic effects of the water caused by earthquake and temperature effects caused by failure of the Spent Fuel Cooling System. This structure was designed by moment distribution methods. The stresses in the walls between the spent fuel pit and fuel transfer canal and those between the spent fuel pit and cask loading area were checked by the ICES STRUDL-II, Volume II computer program to determine the effect of the slot in the walls. Procedures used to design the structural steel framing were based on simple beam and column construction as covered in AISC 'Specification for the Design, Fabrication and Erection of Structural Steel for Buildings,' Part 1, with Type 2 framing connections. Railroad Access Hatch Covers Structural members for the covers were designed as simple beams. Members of the embedded frame were considered as being rigidly supported by concrete. Loads from the embedded frame are transferred to the concrete by embedded anchors. The earthquake forces, specified as follows for design, were determined by dynamic analysis including amplification through the supporting structure. Accelerations for the SSE were used as static loads for determining component and member sizes. After establishing the component and member sizes, a dynamic analysis, using appropriate response spectrum was made of the covers to determine that allowable stresses had not been exceeded. Railroad Access Door The horizontal structural members of the door were designed as simple beams with uniformly distributed loads. The end reactions from these members were then transferred to the door end posts as concentrated loads located between rollers. As a conservative design, it was assumed that one roller was not in contact with the track and that the loading from the two horizontal members with the highest reactions was carried by the two adjacent rollers. The skin plate for the door was designed, without regard to support of the plate from diagonal stiffeners, for the largest open rectangle within the structure. The plate was assumed to be a rectangular diaphragm with fixed edges. 3.8.4-30
WBN The embedded door frame is rigidly supported by concrete. The portions of the frame which form the door track were designed as cantilever members with loading as applied by the door rollers. The structural members of the steel enclosure above the door were designed as simple beams and the hoist supports as cantilevers from the Auxiliary Building wall. Earthquake loads used in design of the door, frame, and track were the loads produced by a SSE having peak accelerations at ground level Elevation 729.0, which is the bottom of the door. Earthquake loads used in design of the hoist supports and enclosure were the loads due to accelerations at the hoist platform, Elevation 773.0, produced by a SSE. These accelerations were determined by dynamic analysis of the Auxiliary Building structure. These accelerations were used as static loads for determining component and member sizes. After establishing the component and member sizes, a dynamic analysis, using appropriate response spectra, was made of the door, embedded frame, door track, and hoisting unit enclosure to determine that allowable stresses had not been exceeded. Manways in the (RHR) Sump Valve Room In the closed position, each door was considered as a structure supported around the periphery. In the open position, each door was considered as a cantilevered structure with the hinges and hinge anchorages being designed for their loading from the door in the open position. Each embedded frame was considered as being rigidly supported by concrete. Loads from the embedded frame are transferred to the concrete by embedded anchors. Earthquake loads used in designing the manways were the forces due to accelerations determined for the sump valve room walls at the center of the manways by dynamic analysis of the Auxiliary Buildings for an OBE or SSE. These forces were used as static loads since the manways are rigid and firmly secured to the walls when closed. Pressure Confining Personnel Doors Structural members for the doors, in the closed position, were designed as simple beams with end reactions carried by the outside members to the frames which were considered as being rigidly supported by concrete. Loads are transferred to the concrete through embedded anchors or bolt anchors. In the open position, the doors were designed as cantilever structures with resultant concentrated loads being used for design of the hinge members. For design, the earthquake loads for the various doors consisted of the loads produced by an OBE or SSE. 3.8.4-31
WBN-1 Earthquake forces due to building accelerations at the elevation of the center of gravity of the various doors were used as static loads for determining door component and member sizes. The building accelerations were determined by dynamic analysis including amplification through the supporting structures. After establishing the component and member sizes, a dynamic analysis, using appropriate response spectra, was made of the doors to determine that allowable stresses had not been exceeded. Fuel Pool Gates The gates are designed for a waterhead load of 25.0 feet imposed from the fuel pool side as measured from the centerline of the horizontal bottom seal to the normal pool level at Elevation 749.13. The gates are constructed of welded corrosion resistant steel. When dewatering the fuel transfer canal inflatable elastomer seals provide a near watertight seal between the skin plate and the pool wall liner face. The gates have been analyzed for the effects of the OBE and SSE for both the operating and stored position. The gates are designed to maintain their sealing and structural integrity during and after an OBE or SSE. Earthquake loading considers simultaneous vertical and horizontal dynamic forces that act on the gates when there is water either on both sides or on the fuel pool side only. The gates are restrained by guides at the top, mid-height, and bottom. When in the storage position, the gates are horizontally restrained by top and bottom guides and vertically supported by hanger brackets. Waste Packaging Structure The Waste Packaging Structure Building is a shear wall structure. The structure is basically a box-like structure. The sloping roof of the structure is a series of 19 pre-cast 20-inch thick panels which span one-way from the south to north walls of the structure. The entire roof section is covered by a 4-inch thick topping slab which is bonded to the pre cast panels. The response of the structure varies with the direction of loading. Load cases which have lateral motion to the north result in shear wall and roof diaphragm action with a triangular soil distribution developing beneath the structure. Loading which produces motion southward brings the Waste Packing Structure Building against the fibrous glass filler material and the Auxiliary Building. Condensate Demineralizer Waste Evaporator Structure This two story structure is designed using the loads, loading combinations, and allowable stresses as given in Tables 3.8.4-1 and 3.8.4-2. The concrete portion is designed in accordance with the ACI 318-71 Building Code and the structural steel portion in accordance with AISC 'Manual of Steel Construction,' Seventh Edition. The building is designed to be supported by a bearing pile foundation, with the piles founded on sound rock. The intermediate floor and roof are supported by interior bearing walls and metal decking spanning between steel beams. 3.8.4-32
WBN 3.8.4.4.2 Diesel Generator Building The structure is designed in accordance with the ACI 318-63 Building Code and is analyzed as a box-type structure assuming all walls fixed at the base slab, Elevation 742.0. The frame is analyzed by the moment distribution methods. Floor Elevation 760.5 and the roof Elevation 773.5 are one-way slabs continuous across interior walls and restrained at exterior walls. All horizontal forces are transmitted through the floor and roof slabs as diaphragms to parallel shear walls and then to the foundation base slab as discussed in Section 3.8.4.4.1 for the Auxiliary Building. The 9-foot 9-inch base slab distributes superstructure loads uniformly to the supporting crushed stone fill and was analyzed as a flat slab. The exterior walls and roof of the building are designed to resist the tornado-generated missiles of Spectrum A in Table 3.5-7. Due to the openings in the exterior walls and floor slab at Elevation 760.5, the building is assumed to depressurize. In the hallway and stairway the glass is assumed to break in the event of a tornado, thereby preventing pressure buildup. A reinforced concrete curb is provided to protect the diesel exhaust stacks from closure due to the impact of tornado-generated missiles. The exhaust stacks extend 24 inches above roof level. The concrete curb is 18 inches thick and extends 12 inches above the exhaust stacks. The fuel oil storage tank vent lines on the roof are encased in concrete to prevent closure due to missile impact. Details of the curbs, exhaust stacks, and fuel oil storage tank vent line encasements are shown on Figures 3.8.4-26 and 3.8.4-33. Missile entry through the air-intake opening in the ceiling over each electrical board room is prevented by the use of steel canopy with barrier protection to intercept missiles. The items discussed above are also listed in Table 3.5-14. Concrete block walls are shown on Figures 3.8.4-24 and 3.8.4-25. All reinforced masonry walls are designed in accordance with ACI 531-79 and NUREG-0800, Section 3.8.4, Appendix A. Diesel Generator Building Doors and Bulkheads Structural members for the doors and bulkheads were designed as simple beams. The skin plates were designed as square or rectangular diaphragms with all edges fixed. 3.8.4-33
WBN Earthquake loads used in designing the doors and bulkheads were the accelerations determined for ground level Elevation 742.0, which is the bottom of the doors, for an SSE. These accelerations were used as static loads for determining component and member sizes. After establishing the component and member sizes, a dynamic analysis was made of the doors and bulkheads. The precast concrete bulkheads (see Figure 3.8.4-33) covering the doors were analyzed for the missile impact loads discussed in Section 3.8.4.1.2. 3.8.4.4.3 Category I Water Tanks and Pipe Tunnels See Section 9.2.7 for a description of the refueling water storage tanks. Pipe Tunnels The pipe tunnels were analyzed using the ICES STRUDL-II Volume I computer program frame analysis and designed in accordance with the provision of the ACI 318-71 Building Code. 3.8.4.4.4 Class 1E Electrical System Manholes The manholes were analyzed using a continuous frame or a series of flat plates, depending upon the boundary conditions created by the duct run openings. The frames were modeled using the computer program STRESS. The concept of joint continuity was utilized with the plate analysis, i.e., joints were designed for the larger moment from adjacent plates. The design of the duct runs assumes bending moments due to earthquake loading are caused by direct imposition of the soil curvature on the duct bank. 3.8.4.4.5 North Steam Valve Room The concrete structure is analyzed as a three-sided open box structure. The 7-foot-thick base slab is designed to span between the foundation walls. The slab is subjected to a pressure loading due to a main steam pipe rupture as well as anchorage loads from restraints located in the slab. The slab was designed using the SAP IV Finite Element computer program. No support from the soil and crushed stone beneath was assumed in the design of the slab. The main steam and feedwater lines exit from the 4-foot-thick west wall where restraints for these lines are anchored. Pipe restraints are also located in the 5-foot thick interior wall in the east end as well as in the 7-foot by 10-foot-deep beam portion of the north wall. The 5-foot interior wall at the east end stiffens the 3-foot-thick east exterior wall. The 2-foot-thick north wall spans horizontally between the stiff complex of end walls and vertically from the base slab to the 7-foot-thick beam portion. The walls are investigated using the SAP IV Finite Element computer programs. 3.8.4-34
WBN Design procedures for the roof steel were based on simple beam construction as covered in AISC "Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings," Part 1 with Type 2 framing connections. The metal decking was attached to a cold formed steel frame which in turn is attached to structural steel with the appropriate number of pressure relief fasteners designed to fail allowing the deck/frame to blow off when the internal pressure at the roof reaches 72 pounds per square foot. The deck/frame is restrained from becoming a missile by using wire rope and clamps which are attached to the main concrete structure. 3.8.4.4.6 Intake Pumping Station and Retaining Walls The box-type structure is analyzed using conventional structural analysis methods. In accordance with ACI 318-71 Code and subsequent addenda, the alternate design method is used in the design of the structure. The base slab was analyzed as a flat plate fixed on four sides for areas within the walls. The overhanging areas of the base slab were analyzed as a cantilever or flat plate fixed on three sides. The other floors were analyzed as flat plates with either three or four sides fixed. The walls were analyzed as one-way span vertically for the first 20 feet, above that the walls were analyzed as flat plates fixed on four sides. The missile barrier walls around the top deck were analyzed as cantilevers. The structure is investigated as a whole to ensure continuity of design. The structure has also been investigated for stability against overturning, floating, and sliding. In addition, the structure is designed to resist the pressure differential during a tornado and to maintain its stability under all credible environmental conditions. The structural adequacy is also checked for missile penetration. Concrete Retaining Walls The concrete retaining walls were analyzed as cantilevers. The walls have also been investigated for stability against overturning and sliding. Sheet Pile Retaining Walls The sheet pile walls are connected by ties to a common concrete "dead man" placed midway between the walls in the earthfill. The ties are steel cables and are anchored in the "dead man" at one end and the sheet piling at the other end. The steel walls are on the inside of the sheet piling and bolted to each pile to transfer the reaction of the pile to the wall. The wall was divided into sections and analyzed as a multibraced wall or cantilever wall depending upon the depth of backfill on the wall. 3.8.4-35
WBN 3.8.4.4.7 Miscellaneous ERCW Structures Slabs and Beams Supporting ERCW Pipes The slab supporting the ERCW pipes was analyzed by the use of McDonnell-Douglas' ICES STRUDL computer program. Support was assumed to be furnished entirely by the bearing piles and the piles were designed for the reaction from the computer analysis. Missile protection is provided by roller compacted concrete above the pipes. The beam encasing the ERCW pipes are analyzed as simple beams with no support from the soil. The encased pipes are in the tension zone of the beam; therefore, the design is for a rectangular beam with no special consideration given to the embedded pipe for flexure or shear. The concrete encasement is designed for missile penetration. Discharge Overflow Structure The discharge overflow structure was analyzed assuming it as a series of flat plates. The concept of joint continuity was utilized with the plate analysis by designing the joints for the larger moment from adjacent plates. Standpipe Structures The standpipe structures consist of a free standing cantilever supported on a flat slab base on in-situ soil. Generally, the structures were considered solid mass concrete and the design was controlled by structural response for missile impact utilizing an elastic analysis. Valve Covers The function of these structures is solely to protect the ERCW valves from tornado missiles; therefore, the design was for missile penetration only. Missile Protection Slabs and Backfill See Section 3.8.4.1.7 3.8.4-36
WBN 3.8.4.4.8 Additional Diesel Generator Building The building is a 96 feet long by 53 feet wide by 32 feet tall (measured from top of base slab) reinforced concrete structure, consisting of a base slab supported by end bearing H-piles, interior floor, roof, and interior and exterior walls. The structure was analyzed as a box-type structure assuming all walls are fixed at the base slab. The building span in the short direction is analyzed using a STRUDL frame program and is designed to withstand all loading conditions assuming a one-way span. In the short direction the interior walls are not considered effective shear walls, but the exterior walls are. Therefore, shear wall and diaphragm deflections are considered in the short direction frame analysis. The building span in the long direction is designed using standard plate theory assuming the interior and exterior walls effectively prevent side sway. The building base slab is a 96 foot long by 53 foot wide by 12 foot thick reinforced concrete slab supported by 154 end-bearing steel H-piles. See Section 3.8.5.5 for additional information on the piles and base slab. The load definitions, load combinations and allowable stresses are as specified in Section 3.8.4.3.2. Base Slab Design The base slab is pile supported. The slab was designed for a uniform live load except where equipment weights dictated a higher value. Equipment loads due to vibration or earthquake acceleration that were transmitted to the slab from anchor bolts were also taken into consideration. In addition, the slab was designed for hydrostatic pressures. The base slab is a rectangular, cast-in-place, reinforced concrete structure with embedded diesel fuel storage tanks and is supported by piles bearing on rock. Roof Slab Design The roof slab was designed for live, seismic, and tornado loads. Floor Slab The floor slab is a poured-in-place reinforced concrete slab designed to carry and transmit the floor loads to the building walls. The slab was designed for a uniform live load. Exterior Walls The building was designed for tornado venting. However, the exterior walls were designed for tornado, wind and seismic loads. 3.8.4-37
WBN Fuel Oil Storage Tanks The steel liner serves no other function except to maintain leak tightness and, therefore, was designed in accordance with ASME Boiler and Pressure Vessel Code, Section VIII, Division I. In addition, the liner was designed to prevent buckling of the steel shell due to the following external loads:
- a. Hydrostatic pressure from underground water.
- b. Shrinkage of the concrete encasement during construction.
- c. Expansion or contraction due to temperature differentials.
For flammable liquids storage requirements, the fuel oil storage tanks meet the requirements of the National Fire Protection Association (NFPA) Code 30, which applies to fuel oil storage tanks supplying underground storage of a Class II liquid (diesel fuel). Equipment Door The equipment door is composed of a structural steel frame and covered on both sides with a steel-skin plate. The removable precast concrete missile barrier bulkheads are placed in front of the equipment doors to provide protection from tornado-generated missiles, which are discussed in Section 3.8.4.1.8. In establishing the required concrete thickness for these missile barriers, no consideration was given to the equipment door. Therefore, these barriers are designed to absorb the full missile impact. Allowable Settlement The building was designed to accommodate a settlement of 2 inches, with a differential settlement of 1-inch over a 96-foot structure length. End-Bearing Steel H-Piles The piles were designed to withstand and transmit to rock the effects of the design loads and conditions. Seismic Analysis The structure was analyzed for the effects of the OBE and the SSE as described in Section 3.7.2.1.1. 3.8.4-38
WBN 3.8.4.5 Structural Acceptance Criteria 3.8.4.5.1 Concrete The Category I structures were proportioned to maintain elastic behavior and stresses within stress allowables when subject to the loading combinations of Section 3.8.4.3. Most Category I structures are essentially low profile box structures with height to base ratios less than 1 and a high factor of safety against sliding or overturning under the most severe loading conditions. Those structures with height to base ratios greater than 1 are designed with adequate factors of safety applied to stability. In addition, all structures are designed to flood or have sufficient weight to prevent flotation under maximum flood conditions. For consideration of sliding, overturning, and floatation of the Additional Diesel Generator Building, see the loading combinations and minimum factor of safety in Table 3.8.4-22. 3.8.4.5.2 Structural and Miscellaneous Steel Structural and miscellaneous steel (including inside containment) and welds are designed in accordance with AISC "Manual of Steel Construction," Seventh Edition, for Case I loading condition so that the stress in the members and connections do not exceed the allowable stress criteria as set forth in the February 1969 AISC "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," as amended through June 12, 1974. For the factor of safety of these allowable stresses with respect to specified minimum yield point of the material used, see Section 1.5 of "Commentary on the Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings." Both specifications and commentary are included in the AISC "Manual of Steel Construction." For Case II loading condition the actual stresses do not exceed the allowable stresses as set forth in Table 3.8.4-2. The allowable stresses for Case II loading have a minimum factor of safety of 1.11 based on the specified minimum yield point of the material used. TVA has generally installed, and will continue to install fillet welds to meet the minimum weld size specifications of Table 1.17.5 of AISC Manual of Steel Construction. Where TVA drawings have specified fillet welds below the minimum sizes specified by AISC, these welds do meet the allowable stress requirements identified above. Weld qualification testing has demonstrated the adequacy of all fillet welds that were installed below minimum AISC specifications. The Additional Diesel Generator Building structural steel was proportioned to meet the applicable codes discussed in Appendix 3.8E and load combinations in Section 3.8.4.3. Structural steel and miscellaneous steel, which is highly restrained and is located in a high temperature environment, is evaluated for effects of thermal loads. 3.8.4-39
WBN 3.8.4.5.3 Miscellaneous Components of the Auxiliary Building Control Room Shield Doors Allowable stresses for all load combinations used for the various parts of the door and dogs are given in Table 3.8.4-3. For normal load conditions the allowable stresses provide a safety factor of 2 to 1 on yield for structural parts and 5 to 1 on ultimate for mechanical parts. For the limiting condition of SSE, stresses do not exceed 0.9 yield. Watertight Equipment Hatch Covers Allowable stresses for all load combinations used for the various parts are given in Table 3.8.4-
- 23. Allowable stresses for normal loading combinations are based on the AISC specification (see Section 3.8.4.2 and Table 3.8.4-23). For limiting conditions, such as SSE, tornado, and flood, stresses do not exceed 0.9 yield.
Railway Access Hatch Covers Allowable stresses for all load combinations used for the various parts are given in Table 3.8.4-4. For normal load conditions, the allowable stresses provide safety factors of 2 to 1 on yield for structural parts and 5 to 1 on ultimate for mechanical parts. For limiting conditions, such as an OBE or SSE, stresses do not exceed 0.9 yield. Railroad Access Door Allowable stresses for all load combinations used for the various parts of the door, embedded frame, and hoist enclosure are given in Table 3.8.4-5. For normal load conditions the allowable stresses provide a safety factor of 2 to 1 on yield for structural parts and 5 to 1 on ultimate for mechanical parts. For limiting conditions such as an OBE or SSE and hoist stall, stresses do not exceed 0.9 yield. Manways in RHR Sump Valve Room Allowable stresses for all load combinations used for the various parts are given in Table 3.8.4-6. For limiting conditions, such as a SSE, stresses do not exceed 0.9 yield. 3.8.4-40
WBN Pressure Confining Personnel Doors Allowable stresses for all load combinations used for the various parts are given in Table 3.8.4-7. For normal load conditions, the allowable stresses provide safety factors of 2 to 1 on yield on structural parts and 5 to 1 on ultimate for mechanical parts. For limiting conditions, such as an SSE, flood, and tornado loadings, stresses do not exceed 0.9 yield. Fuel Pool Gates Allowable stresses for all load combinations used for the gates are given in Table 3.8.4-21. For normal load conditions the allowable stresses do not exceed 0.6 of yield. For limiting conditions, such as the SSE, the stresses do not exceed 0.90 of yield. 3.8.4.5.4 Intake Pumping Station Traveling Water Screens Allowable stresses for all load combinations used for the various parts are given in Table 3.8.4-
- 11. For normal load conditions, the allowable stresses provide safety factors of 1.79 (Fy/0.56 Fy) to 1 on yield for structural parts and 5 to 1 on ultimate for mechanical parts. For limiting conditions, such as a safe shutdown earthquake, stresses do not exceed 0.9 yield.
3.8.4.5.5 Diesel Generator Building Doors and Bulkheads Load combinations and allowable stresses for all combinations are given in Table 3.8.4-13. For missile impact, yield point of material will be exceeded and the member practically deform. For normal load condition, the allowable stresses provide safety factors of 1.67 (Fy/0.6 Fy) to 1 on yield for structural parts and 5 to 1 on ultimate for mechanical parts. For limiting conditions, except for missile impact, stresses do not exceed 0.9 yield. 3.8.4.5.6 Additional Diesel Generator Building Missile Barriers Design of missile barriers for the Additional Diesel Generator Building is discussed in Section 3.5.3.1. 3.8.4.6 Materials, Quality Control, and Special Construction Techniques General See Section 3.8.1.6. 3.8.4.6.1 Materials See Section 3.8.1.6.1. For the Additional Diesel Generator Building the following materials were used: 3.8.4-41
WBN Structural Steel Rolled shapes, plates, and bars meet Specification ASTM A 36. Fabricated high-strength steel meets Specification ASTM A 572 and bolting meets Specification ASTM A 325 or A 490. Anchor bolts meet ASTM A 307 or A 36 steel. Reinforcing Steel The yield strength of reinforcing steel used in the building is 60,000 psi (ASTM A 615, grade 60) or greater. Concrete The compressive strength of concrete is 3000 psi or greater. 3.8.4.6.2 Quality Control Concrete production and testing are discussed in Section 3.8.1.6.2. In addition to the 4,000-psi-at-28-days mix discussed in Section 3.8.1.6.2, a 3,000-psi-at-28-days mix, a 3,000-psi-at-90-days mix, a 5,000-psi-at-28-days mix, and a 4,000-psi-at-90-days mix were used. Some concrete did not meet specification requirements. This was evaluated and documented in the report CEB-86-19-C "Concrete Quality Evaluation."[2] Results have been documented in affected calculation packages and drawings. Testing of reinforcing steel and cadweld splices is discussed in Section 3.8.1.6.2. The control room shield doors, watertight equipment hatch covers, railway access hatch covers, railroad access doors, equipment hatch doors and sleeves, manways in the RHR sump valve room, and the pressure confining personnel doors were designed and erected by TVA in accordance with TVA's quality assurance program. Design and fabrication by the contractor were in accordance with the contractor's quality assurance program which was reviewed and approved by TVA's design engineers. The contractor's quality assurance program covers the criteria in Appendix B of 10 CFR 50. Fabrication procedures such as welding and nondestructive testing were included in appendices to the contractor's quality assurance program. ASTM standards were used for all material specifications and certified mill test reports were provided by the contractor for materials used for load-carrying members. The fuel pool gates were designed and procured before quality assurance requirements were imposed. An evaluation was conducted to verify that the gates were equivalent to gates that would have been designed and fabricated under a quality assurance program. 3.8.4-42
WBN 3.8.4.6.3 Special Construction Techniques No special construction techniques were used, except for the fuel oil storage tanks in the Additional Diesel Generator Building. For these tanks joint welding procedures used in fabrication of the steel liner were qualified in accordance with ASME Boiler and Pressure Vessel Code, Section IX, prior to use by TVA or the fabricator. 3.8.4.7 Testing and Inservice Surveillance Requirements Testing for the steel liners in the fuel oil storage tanks for the Additional Diesel Generator Building was accomplished by subjecting them to a standard hydrostatic test in accordance with ASME, Section VIII. 3.8.4.7.1 Concrete and Structural Steel Portions of Structures A program to monitor the settlement of other Category I structures is as shown in Figures 3.8.4-66 and 3.8.4-67. 3.8.4.7.2 Miscellaneous Components of Auxiliary-Control Building Control Room Shield Doors After erection and adjustment the doors were inspected for proper operation of the dogs and free movement on the trolleys. After the initial inspection, periodic visual inspections of the doors are to be made. Parts inspected during these visual inspections are to include connections to trolleys, structural members for paint deterioration, and dogs. Watertight Equipment Hatch Covers After initial inspection, periodic visual inspections of the covers are to be made. A visual inspection is made of all screws to see that they are securely tightened and that none are missing. The painted inscriptions on the covers are inspected for any deterioration. In the event that the hatch covers are removed, an inspection is made of the gaskets to ensure that they are clean and free of any damage or deterioration which would prevent their forming a proper seal. The embedded frames are inspected to ensure that the mating surfaces are clean and free of foreign material before the covers are reinstalled. 3.8.4-43
WBN Railway Access Hatch Covers After the initial inspection, periodic visual inspections of the covers are made. Parts inspected during the visual inspection are to include all bolted connections, structural members for paint deterioration, limit switches, and rubber seals. The seals are carefully, inspected for cracks, blemishes, or any other indications of deterioration of the rubber and for properly seating at the sealing surfaces. Railway Access Door Prior to shipment of the door from the contractor's plant, the splice welds in the skin plate of the door and welds among the periphery of the skin plate and structural members were magnetic particle tested. After completion of the initial tests and inspection, periodic visual inspections of the door and its parts are made. Parts inspected are to include all bolted connections, limit switches, door tracks, and rollers. Painting is to be inspected for evidence of deterioration, and the seals are carefully inspected for cracks, blemishes, or any other indications of deterioration of the rubber. Pressure Confining Personnel Doors After the initial inspection, periodic visual inspections of the doors are made. Parts inspected during these visual inspections include all bolted connections, structural members for paint deterioration, latching or dogging mechanisms and limit switches for physical condition, and the seals. The seals are carefully inspected for cracks, blemishes, or any other indications or deterioration and for proper seating at the sealing surfaces. Fuel Pool Gates After initial inspection, periodic visual inspections of the gates are made. The seals are carefully inspected for cracks, blemishes, or any other indications of deterioration. 3.8.4.7.3 Miscellaneous Components of the Intake Pumping Station Traveling Water Screens After the initial inspection and testing, the screens are inspected at periodic intervals. Parts inspected include drive components, carrier chains, baskets including the wire panels, spray pipes, spray nozzles, main frames, lights, and lubricating devices. 3.8.4-44
WBN Watertight Doors After the initial inspection, periodic visual inspections of the doors are made. Parts inspected during these visual inspections include the bolted connections, structural members for paint deterioration, latching or dogging mechanisms and limit switches for physical condition, and the seals. The seals are carefully inspected for cracks, blemishes, or any other indications or deterioration and for proper seating at the sealing surfaces. REFERENCES
- 1. TVA drawing series 46W454 "Architectural Door and Hardware Schedule."
- 2. TVA Civil Engineering Branch Report Number CEB-86-19-C, "Concrete Quality Evaluation."
- 3. Portland Cement Association publication, T18-4, "Analysis of Small Reinforced Concrete Buildings for Earthquake Forces," pp. 30-32.
3.8.4-45
WBN TABLE 3.8.4-1 (SHEET 1 of 4) AUXILIARY-CONTROL BUILDING CONCRETE STRUCTURE LOADS, LOADING COMBINATIONS AND ALLOWABLE STRESSES I. Loads The following terms are used in the load combination equations. C = Construction condition. C' = Crane load, including wind on crane. D = Dead load of structure and equipment plus any other permanent load contributing stress, such as soil pressure. Hydrostatic pressure from ground water Elevation 710, exterior walls; Elevation 726, uplift. D' = D + hydrostatic pressure from groundwater Elevation 724.4. E = Operating Basis Earthquake E' = Safe Shutdown Earthquake H = Spent fuel pit hydrostatic pressure. Worst condition of the following except as noted:
- 1. Normal water level in pit, cask loading area and canal.
- 2. Canal empty of water. Normal water level in other areas.
- 3. Cask loading area empty of water. Normal water level in other areas.
Considered for Case I load combinations only. L = Live load. For live load on slabs, see Figure 3.8.4-9. P = Accidental drop of fuel cask on walls of cask loading area. Ta = Accidental increase in temperature of water in pit to 212°F in 8 hours. Temperature inside building 60°F. TN = Normal temperature of water in fuel pit and canal 120°F. Temperature inside building 60°F.
WBN TABLE 3.8.4-1 (SHEET 2 of 4) AUXILIARY-CONTROL BUILDING CONCRETE STRUCTURE LOADS, LOADING COMBINATIONS AND ALLOWABLE STRESSES (Cont'd) W = Wind load, see Section 3.3. Wt = Tornado, see Section 3.3. Pa = Pressure from main steam break. Ra = Pipe reaction from thermal effects of main steam break. Ta = Thermal effects from main steam break. Yr = Pipe anchor force due to jet from pipe break. Yj = Jet force from pipe break. Ym = Missile impact froce from pipe whip. II. Load Combinations and Allowable Stresses Auxiliary-Control Building Allowable Load combinations WSD Stresses Case I = D+L Normal (ACI 318-63 or 318-71) Case Ia = D'+L 1.35 x normal Case Ib = D+L+W+C 1.33 x normal Case II = D+L+E fc = normal (ACI 318-63 or 318-71) fs = 0.50 fy Case III = D+L+E' *fc = 0.75 f'c fs = 0.90 fy Case IV = D+L+Wt *fc = 0.75 f'c fs = 0.90 fy
WBN TABLE 3.8.4-1 (SHEET 3 of 4) AUXILIARY-CONTROL BUILDING CONCRETE STRUCTURE LOADS, LOADING COMBINATIONS AND ALLOWABLE STRESSES (Cont'd) Where main steam lines pass through the Auxiliary-Control Building at the south main steam valve room, the following factored load combinations were considered in addition to those listed above: Allowable USD Load Combinations WSD Stresses Load Factors Case VI = D+L+Pa *fc = .75f'c 1.0D+1.0L+1.5Pa fs = .9fy Case VII = D+L+Pa+1.0 *fc = .75f'c 1.0D+1.0L+1.25Pa+ Yr+Yj+Ym)+1.0E fs = .9fy 1.0(Yr+Yj+Ym)+1.25E Case VIII = D+L+1.0Pa+1.0 *fc = .75f'c Yr+Yj+Ym)+1.0E' fs = .9fy 1.0(D+L+Pa+Yr+Yj+Ym +E')
- Concrete stresses other than flexure = 1.67 x normal Where the main steam lines pass under the elevation 755.0 floor slab of the Control Building, vital structural elements in that area were designed for cases I through IV and the case IX listed below.
Allowable WSD Stresses Case IX = D+E'+J* fc = 1.67 (Normal Concrete) fs = 0.9 fy
- J is a jet load of 360 kips spread over 50 ft2 Material Properties Concrete Slabs and walls f'c = 3000 or 4000 psi Columns f'c = 4000 psi Concrete weight w = 145 pcf Reinforcing steel fy = 60,000 psi (ASTM A615. Grade 60)
WBN TABLE 3.8.4-1 (SHEET 4 of 4) AUXILIARY-CONTROL BUILDING CONCRETE STRUCTURE LOADS, LOADING COMBINATIONS AND ALLOWABLE STRESSES (Cont'd) Auxiliary Building Spent Fuel Pit Allowable Load Combinations WSD Stresses Case I = D+H Normal (ACI 318-63)
= D+H+TN Normal (ACI 318-63)
Case II = D+H+E fc = normal (ACI 318-63) fs = 0.50 fy
= D+H+E+TN fc = normal (ACI 318-63) fs = 0.50 fy Case III = D+H+E' *fc = 0.75 f'c fs = 0.90 fy
= D+H+E'+TN *fc = 0.75 f'c fs = 0.90 fy Case IV = D+H+Ta *fc = 0.75 f'c fs = 0.90 fy Case IVa = D+H+P *fc = 0.75 f'c fs = 0.90 fy
- Concrete stresses other than flexure = 1.67 x normal.
Material Properties (see above) Auxiliary Building Concrete Structure Earth Values Angle of internal friction = 32° Angle of friction between fill and structure F = 16° Unit weight of fill Surcharge w = 120 psf Dry Saturated w = 65 psf Surcharge A1 and A15 line walls 1730 psf Others 200 psf
WBN TABLE 3.8.4-2 (Sheet 1 of 4) AUXILIARY-CONTROL BUILDING STRUCTURAL STEEL LOADS, LOADING CONDITIONS AND ALLOWABLE STRESSES Control Building Portion
- 1. Live Loads (LL)
- a. Elevation 755.0 - 400 psf (to include cable trays, ducts, walls, and electrical boards)
- b. Elevation 741.0 - 100 psf when seismic loads (E and E) are not present, 10 psf when seismic loads are present
- c. Elevation 729.0 - 100 psf
- 2. Dead Loads (DL)
- a. 8-inch concrete brick wall - 100 psf
- b. 1-1/2-inch steel grating - 12 psf
- c. Concrete - 12.5 psf per inch thickness
- d. Steel framing - 15 psf
- e. Piping - varies
- 3. Tornado (TOR)
- a. Elevation 729.0 - 3.0 psi (between column lines C-1 to C-3 and C-11 to C-13).
Auxiliary Building Portion
- 1. Live Loads (LL) - The following loads shall be used unless shown otherwise on Figure 3.8.4-9, Concrete Floor Design Data.
- a. Roof (Refueling Area) - 50 psf or equipment
- b. Roof (Main steam valve room) - 50 psf when seismic loads are not present and 5 psf when seismic loads are present, or equipment
- c. Floor (Main steam valve room) - 100 psf or equipment
- d. Roof (CDWE Building) - 300 psf or equipment
- e. Floor (CDWE Building) - 300 psf or equipment
- f. Construction Load - 20 psf
- g. Miscellaneous live load - 30 psf
- 2. Dead Loads (DL)
- a. Concrete - 12.5 psf per inch thickness
- b. Steel roof decking - 4 psf
- c. Steel roof framing - 30 psf
- d. Steel floor framing - 15 psf
- 3. Tornado (TOR)
- a. Velocity - 360 mph
WBN TABLE 3.8.4-2 (Sheet 2 of 4) AUXILIARY-CONTROL BUILDING STRUCTURAL STEEL LOADS, LOADING CONDITIONS AND ALLOWABLE STRESSES (Cont'd) Auxiliary-Control Building Seismic Loads
- a. Operating Basis Earthquake (OBE) maximum ground acceleration Horizontal 0.09g Vertical 0.06g
- b. Safe Shutdown Earthquake (SSE) maximum ground acceleration Horizontal 0.18g Vertical 0.12g Shear on Compression Loading Tension on Gross on Gross Concrete Condition Net Section Section Section Bending Bearing Case I 0.60 FY 0.40 FY See Note 1 0.66 FY to 0.25 f'c DL + LL + OBE -0.60 FY Case II 0.90 FY 0.9 FY See Note 2 0.90 FY .595 f'c DL + LL + SSE 3 See Note 3 Case III 0.90 FY 0.9 FY See Note 2 0.90 FY .595 f'c DL + LL + TOR v 3 See Note 3 Note 1 - Varies with slenderness ratio, see AISC "Manual of Steel Construction," 7th Edition, Table 1-36, Page 5-84.
Note 2 - Varies with slenderness ratio: Main and secondary members, where KL/r Cc: (KL / r ) 2 Fa = 0.9FY 1 (A) 2C c2
WBN TABLE 3.8.4-2 (Sheet 3 of 4) AUXILIARY-CONTROL BUILDING STRUCTURAL STEEL LOADS, LOADING CONDITIONS AND ALLOWABLE STRESSES (Cont'd) Main members, where Cc < KL/r < 200: Fa = 0.92E (B) (KL/r)2 Secondary members, where 120 < L/r < 200: Fas = Fa [by Formula (A) or (B)] 1.6 - L/200r Where: 2 2 E Cc = FY E = Modulus of elasticity of steel Fa = Axial compressive stress permitted in the absence of bending moment (kips per square inch) Fas = Axial compressive stress, permitted in the absence of bending moment, for bracing and other secondary members (kips per square inch) FY = Specified minimum yield stress of material (kips per square inch) f'c = Compressive strength of concrete K = Effective length factor L = Actual unbraced length (inches) r = Governing radius of gyration (inches) Material Properties Steel Properties Cc = 126.1 E = 29,000,000 psi Fy = 36,000 psi
WBN TABLE 3.8.4-2 (Sheet 4 of 4) AUXILIARY-CONTROL BUILDING STRUCTURAL STEEL LOADS, LOADING CONDITIONS AND ALLOWABLE STRESSES (Cont'd) Note 3- When the supporting surface is wider on all sides than the loaded area, the permissible bearing stress on the loaded area may be multiplied by -~F A-2 / A , 1 but not more than 2. 1 Where: A1 = Loaded area A2 = Maximum area of the portion of the supporting surface that is geometrically similar to and concentric with the loaded area.
WBN TABLE 3.8.4-3 CONTROL ROOM SHIELD DOORS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Door and Jamb Shield Assemblies Structural Parts Allowable Stresses (psi) No. Load Combinations Tension Compression(2) Shear Doors Open or Closed I Dead 0.50Fy 0.47Fy 0.33Fy II Dead + OBE(1) 0.60Fy 0.60Fy 0.40Fy III Dead + SSE(1) 0.90Fy 0.90Fy 0.60Fy Mechanical Parts Allowable Stresses (psi) No. Load Combinations Tension Compression(2) Shear Doors Open or Closed I Dead Ultimate 5 Ultimate 5 Ultimate 7.5 II Dead + OBE(1) 0.6FY 0.6Fy 0.4Fy III Dead + SSE(1) 0.9Fy 0.9Fy 0.6Fy Notes: (1) Acts in any one horizontal direction only at any given time and acts in vertical and horizontal directions simultaneously. (2) The value given for allowable compression stress is the maximum value permitted when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls. Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or 2 E Kl Fcr = when > Cc Kl 2 r 2 r
WBN TABLE 3.8.4-4 (SHEET 1 of 2) AUXILIARY BUILDING RAILROAD ACCESS HATCH COVERS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Cover Structure and Embedded Frame Allowable Stresses (psi) Shear No. Load Combinations Tension Compression(1) Covers Closed I Dead load plus live load at 100 lb/ft2 0.50Fy 0.47Fy 0.33Fy II Dead load plus live load at 100 lb/ft2 plus 0.60Fy 0.60Fy 0.40Fy OBE III Dead load plus live load at 100 lb/ft2 plus 0.90Fy 0.90Fy 0.60Fy SSE Covers Open IV Dead load plus hoist pull 0.50Fy 0.47Fy 0.33Fy V Dead load plus hoist pull plus OBE 0.60Fy 0.60Fy 0.40Fy VI Dead load plus hoist pull plus SSE 0.90Fy 0.90Fy 0.60Fy Mechanical Parts on Covers and Frame Allowable Stresses (psi) No. Load Combinations Tension and Compression(1) Shear Covers Closed I Dead load plus live load at 100 lb/ft2 Ult 2 x Ult 5 15 II Dead load plus live load of 100 lb/ft2 plus 0.6Fy 0.4Fy OBE III Dead load plus live load at 100 lb/ft2 plus 0.9Fy 0.6Fy SSE Covers Open IV Dead load plus hoist pull Ult 2 x Ult 5 15 V Dead plus live load of 100 lb/ft2 plus OBE 0.6Fy 0.4Fy VI Dead load plus hoist pull plus SSE 0.9Fy 0.6Fy
WBN TABLE 3.8.4-4 (Sheet 2 of 2) AUXILIARY BUILDING RAILROAD ACCESS HATCH COVERS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Hoist Unit Supports Allowable Stresses(psi) No. Load Combinations Tension Compression(1) Shear Hatch Opening I Dead load Hoist pull 18,000 17,000 12,000 II Dead load Stall 32,400 32,400 21,600 Other Mechanical Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression(1) Shear Covers Open I Dead load Hoist pull Ult 2 x Ult 5 15 II Dead load Stall 0.9Fy 2/3 x 0.9Fy Notes: (1) The value given for allowable compression stress is the maximum value, Fcr, permitted when buckling does not control. The critical buckling stress shall be used in place of Fy when buckling controls. Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or 2 E Kl Fcr = when > Cc Kl 2 r 2 r
WBN TABLE 3.8.4-5 (SHEET 1 of 2) RAILROAD ACCESS DOOR LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Door, Embedded Frame and Door Track Allowable Stresses (psi) No. Load Combinations Tension Compression2 Shear Door Closed I Dead load plus windload at 10 0.50Fy 0.47Fy 0.33Fy lb/ft2 II Dead load plus windload at 30 0.90Fy 0.90Fy 0.60Fy lb/ft2 III Dead load plus windload at 10 0.60Fy 0.60Fy 0.40Fy lb/ft2 plus OBE IV Dead load plus windload at 10 0.90Fy 0.90Fy 0.60Fy lb/ft2 plus SSE Door Open V Dead load plus hoist pull 0.50Fy 0.47Fy 0.33Fy VI Dead load plus hoist pull 0.60Fy 0.60Fy 0.40Fy plus OBE VII Dead load plus hoist pull plus 0.90Fy 0.90Fy 0.60Fy SSE Hoist Unit & Enclosure Allowable Stresses (psi) No. Load Combinations Tension Compression2 Shear I Dead load plus hoist pull 0.50Fy 0.47Fy 0.33Fy II Dead load plus stall 0.90Fy 0.90Fy 0.60Fy III Dead load plus hoist stall plus 0.60Fy 0.60Fy 0.40Fy OBE IV Dead load plus hoist pull plus 0.90Fy 0.90Fy 0.60Fy SSE
WBN TABLE 3.8.4-5 (SHEET 2 of 2) RAILROAD ACCESS DOOR LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES (Cont'd) Mechanical Parts on Door Allowable Stresses (psi) No. Load Combinations Tension and Compression(2) Shear Door Open I Dead load plus windload at 10 lb/ft2 Ult 2x Ult 5 15 II Dead load plus windload a 10 lb/ft2 plus 0.6Fy 0.4Fy OBE III Dead load plus windload at 10 lb/ft2 plus 0.9Fy 0.6Fy SSE(1) Other Mechanical Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression(2) Shear Door Open I Dead load Ult 2 x Ult Hoist pull 5 15 II Dead load Stall 0.9Fy 0.6Fy NOTE: (1) Acts in one horizontal direction only at any given time and acts in the horizontal and vertical directions simultaneously. (2) The value indicated for the allowable compression stresses is the maximum value permitted when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls. Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or 2E Kl Fcr = when > Cc 2 2 Kl r r
WBN TABLE 3.8.4-6 MANWAYS IN RHR SUMP VALVE ROOM LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Structural Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression(2) Shear Manway Closed I Dead load plus OBE(1) 0.6FY 0.4FY II Dead load plus SSE(1) 0.9FY 0.6FY III Dead load plus 19 psi from outside 0.9FY 0.6FY Manway Open IV Dead load plus OBE(1) 0.6FY 0.4FY V Dead load plus SSE(1) 0.9FY 0.6FY Mechanical Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression(2) Shear Manway Closed I Dead load plus OBE(1) 0.6FY 0.4FY II Dead load plus SSE(1) 0.9FY 0.6FY III Dead load plus 19 psi from outside 0.9FY 0.6FY Manway Open IV Dead load plus OBE(1) 0.6FY 0.4FY V Dead load plus SSE(1) 0.9FY 0.6FY NOTES: (1) Acts in one horizontal direction only at any given time and acts in vertical and horizontal directions simultaneously. (2) The values given for allowable compression stress is the maximum value permitted when buckling does not control. The critical buckling stress, Fcr, shall be used in place of FY when buckling controls. Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or 2 E Kl Fcr = when > Cc Kl 2 r 2 r
WBN-3 TABLE 3.8.4-7 (Sheet 1 of 5) PRESSURE CONFINING PERSONNEL DOORS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES1 (All Doors as shown in Table 3.8.4-7a except A55, A57, C20, C26, A101, A105, A216, and A217) Structural Parts Allowable Stresses (psi) No. Load Combinations Tension Compression(2) Shear Doors Open or Closed I DL + Load from Door Closers 0.50 Fy 0.47 Fy 0.33 Fy II DL + OBE + Load from Door Closers 0.60 Fy 0.60 Fy 0.40 Fy DL + SSE + Load from Door Closers 0.90 Fy 0.90 Fy 0.60 Fy Doors Closed III3 DL + 3-psi pressure 0.90 Fy 0.90 Fy 0.60 Fy (bidirectional where applicable) IV4 DL + OBE + 2-psi toward annulus 0.60 Fy 0.60 Fy 0.40 Fy DL + SSE + 2-psi toward annulus 0.90 Fy 0.90 Fy 0.60 Fy V5 DL + 3 inches of water pressure 0.50 Fy 0.47 Fy 0.33 Fy on either side of door VI6 DL + Flood to elevation 739.7 0.90 Fy 0.90 Fy 0.60 Fy
- 1. Thermal load effects are insignificant and hence need not be considered in the design of doors.
- 2. The values indicated for the allowable compression stresses are the maximum values permitted, when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls.
2 Kl r Kl Fcr = FY 1 - 2 when < Cc 1 2 Cc r or 2 E Kl Fcr = 2 when > Cc 2 Kl r r
- 3. Applies to all doors except A56, A60, A64, A65, A77, A78, A94, A99, A111, A117, A118, A122, A123, A125, A130, A132, A133, A151, A152, A159, A160, A161, A162, A183, A192, A204, A206, A207, A208, A209, A212, A213, A214, A215, C36, C37, C49, C50, C53, C54, C60, DE1, DE4 and DE5.
- 4. Applies to doors A64 and A77 only.
- 5. For doors A56, A60, A111, A122, A123, A125, A130, A132, A133, A151, A152, A159, A160, A161, A162, A183, A192, A204, A206, A207, A208, A209, A212, A213, A214, A215, DE1, DE4, and DE5, the load combination is:
DL + 1/2" water pressure on either side of door. For doors C36, C37, C49, C50, C53, C54 and C60, the load combination is: DL + 1/8" water pressure on either side of door.
- 6. Applies to doors A65, A78, A94, and A99.
WBN-3 TABLE 3.8.4-7 (Sheet 2 of 5) PRESSURE CONFINING PERSONNEL DOORS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES1 (Cont'd) (All Doors as shown in Table 3.8.4-7a except A55, A57, C20, C26, A101, A105, and A216) Mechanical Parts Allowable Stresses (psi) Tension and No. Load Combinations Shear Compression2 Doors Open or Closed I DL + Load from door closers Fu/5 2 Fu/15 DL + OBE + Load from door closers 0.60 Fy 0.40 Fy II DL + SSE + Load from door closers 0.90 Fy 0.60 Fy Doors Closed DL + 3-psi pressure III3 0.90 Fy 0.60 Fy (bidirectional where applicable) DL + OBE + 2-psi toward annulus 0.60 Fy 0.40 Fy IV4 DL + SSE + 2-psi toward annulus 0.90 Fy 0.60 Fy DL + 3 inches of water pressure V5 Fu/5 2 Fu/15 on either side of door VI6 DL + Flood to elevation 739.7 0.90 Fy 0.60 Fy
- 1. Thermal load effects are insignificant and hence need not be considered in the design of doors.
- 2. The values indicated for the allowable compression stresses are the maximum values permitted, when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls.
2 Kl r Kl Fcr = FY 1 - 2 when < Cc 3 2 Cc r or 2 E Kl 4 Fcr = 2 when > Cc Kl r r
- 3. Applies to all doors except A64, A65, A77, A78, A56, A60, A111, A117, A118, A122, A125, A130, A133, A151, A160, A162, A183, A192, A206, A207, A208, A209, A212, A213, C37, C49, C50, C53, C60, DE1, DE4 and DE5.
- 4. Applies to doors A64, A65, A77 and A78.
- 5. For doors A56, A60, A65, A94, A111, A113, A114, A122, A123, A125, A130, A132, A133, A151, A152, A159, A160, A161, A162, A183, A192, A206, A207, A208, A209, A212, A213, A214, A215, DE1, DE4, and DE5, the load combination is:
DL + 1/2" water pressure on either side of door. For doors C36, C37, C49, C50, C53, C54, and C60, the load combination is: DL + 1/8" water pressure on either side of door.
- 6. Applies to door A65 and A78 only.
WBN-3 TABLE 3.8.4-7 (Sheet 3 of 5) PRESSURE CONFINING PERSONNEL DOORS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES1 (Cont'd) (Doors A55, A57, C20, C26, A101, and A105) Structural Parts Allowable Stresses (psi) No. Load Combinations Tension Compression2 Shear Doors Open I DL + Load from Door Closers 0.50 Fy 0.47 Fy 0.33 Fy II DL + OBE + Load from Door Closers 0.60 Fy 0.60 Fy 0.40 Fy DL + SSE + Load from Door Closers 0.90 Fy 0.90 Fy 0.60 Fy Doors Closed III3 DL + CCWS Flood + OBE + 3-psi pressure 0.60 Fy 0.60 Fy 0.40 Fy (bidirectional where applicable) IV3 DL + CCWS flood + SSE + 3-psi pressure 0.90 Fy 0.90 Fy 0.60 Fy (bidirectional where applicable) V4 DL + OBE + Pressure from valve rooms 0.60 Fy 0.60 Fy 0.40 Fy DL + SSE + Pressure from valve rooms 0.90 Fy 0.90 Fy 0.60 Fy
- 1. Thermal load effects are insignificant and hence need not be considered in the design of doors.
- 2. The values indicated for the allowable compression stresses are the maximum values permitted, when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls.
2 Kl r Kl Fcr = FY 1 - 2 when < Cc 5 2 Cc r or 2 E Kl Fcr = 2 when > Cc 6 Kl r r
- 3. The CCWS flood condition does not apply to doors A101 and A105, and differential pressure load due to tornado need not be considered simultaneously with seismic load.
- 4. Applies to doors A101 and A105 only.
WBN-3 TABLE 3.8.4-7 (Sheet 4 of 5) PRESSURE CONFINING PERSONNEL DOORS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES1 (Cont'd) (Doors A55, A57, C20, C26, A101, and A105) Mechanical Parts Allowable Stresses (psi) No. Load Combinations Tension and Shear Compression2 Doors Open I DL + Load from door Fu/5 2 Fu/15 closers II DL + OBE + Load from door closers 0.60 Fy 0.40 Fy DL + SSE + Load from door closers 0.90 Fy 0.60 Fy Doors Closed III3 DL + CCWS flood + 3-psi pressure 0.90 Fy 0.60 Fy (bidirectional where applicable) IV4 DL + OBE + Pressure from valve room 0.60 Fy 0.40 Fy DL + SSE + Pressure from valve room 0.90 Fy 0.60 Fy
- 1. Thermal Load effects are insignificant and hence need not be considered in the design of doors.
- 2. The values indicated for the allowable compression stresses is the maximum value permitted, when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls.
2 Kl r Kl Fcr = FY 1 - 2 when < Cc 7 2 Cc r or E Kl 2 Fcr = 2 when > Cc 8 Kl r r
- 3. The CCWS flood condition does not apply to doors A101 and A105, and differential pressure load due to tornado need not be considered simultaneously with seismic load.
- 4. Applies to doors A101 and A105 only.
WBN-3 TABLE 3.8.4-7 (Sheet 5 of 5) PRESSURE CONFINING PERSONNEL DOORS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES1 (Cont'd) (Door A216 and A217) Structural Parts Allowable Stresses (psi) No. Load Combinations Tension Compression2 Shear I DL + P 0.50 Fy 0.47 Fy 0.33 Fy II DL + P + OBE 0.60 Fy 0.60 Fy 0.40 Fy III DL + P + SSE 0.90 Fy 0.90 Fy 0.60 Fy Mechanical Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression2 Shear I DL + P Fu/5 2 Fu/15 II DL + P + OBE 0.60 Fy 0.40 Fy III DL + P + SSE 0.90 Fy 0.60 Fy DL - Stresses generated by dead loads and door closer loads. P - Stresses generated by a pressure differential of 1/2 inch of water acting to open doors.
- 1. Thermal Load effects are insignificant and hence need not be considered in the design of doors.
- 2. The values indicated for the allowable compression stresses is the maximum value permitted, when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls.
2 Kl r Kl Fcr = FY 1 - 2 when < Cc 9 2 Cc r or 2 E Kl Fcr = 2 when > Cc 10 Kl r r
WBN-3 Table 3.8.4-7a LIST OF PERSONNEL ACCESS DOORS IN AUXILIARY/CONTROL BUILDING A55, A57, A64, A65, A78,A94, A95, A96, A101, A105, A113, A114, A115, A123, A132, A152, A153, A154, A155, A156, A157, A158, A159, A161, A164, A165, A173, A184, A191, A214, A215, A216, A217 C20, C26, C29, C34, C36, C54 A111, A117, A125, A130, A151, A160, A162, A183, A192, A206, A207, A208, A209, A212, A213 C37, C49, C50, C53, C60 DE1, DE4, DE5 A56, A60, A122, A133
WBN TABLE 3.8.4-7b ABSCE AIR LOCK BOUNDARY DOORS The following pairs of doors function as ABSCE air lock boundary doors. These doors become part of the ABSCE boundary when the associated airlock boundary door (or damper) is open. Primary ABSCE Boundary Door Associated (Extended Air lock Boundary Door A55 A60 A57 A56 A65 A641 A112 (Railroad access door) A111, A113, A114 A123 A122 A125 A101 DE1 DE4, DE52 A130 A105 A132 A133 A206 A207 A208 A209 A214 A192 A215 A183 NOTES:
- 1. Doors A64 and A65 are interlocked to protect the EGTS (Annulus) pressure boundary. A64 is not an ABSCE boundary component.
- 2. These doors are not required to be interlocked because of the infrequent egress/ingress through them during an ABI.
WBN TABLE 3.8.4-8 (SHEET 1 of 2) INTAKE PUMPING STATION LOADING CASES, ALLOWABLE STRESSES, FACTORS OF SAFETY, AND MATERIAL PROPERTIES WSD Calculated Case Description Allowable Stresses Factors of Safety Overturning Sliding I Reservoir level at El. 690.0, Normal per ACI 318-71 2.27 3.52 operating loads, earthfill, one pump well unwatered (1/2 of structure) II Reservoir level at El. 713.0, Normal Concrete per 1.33 1.39 operating loads, earthfill, ACI fs = 0.5 fy Operating Bases Earthquake (OBE) IIa Reservoir level at El. 732.1 1.35 x Normal per ACI 1.36 2.18 (assuming upstream dam failure), operating loads, earthfill, OBE both pump wells full III Reservoir level at El. 724.4, 1.35 x Normal per ACI N.A. N.A. operating loads, earthfill both pump wells full IV Reservoir level at El. 675.0, 1.67 x Normal per ACI 2.84 2.96 operating loads, earthfill,both fs = 0.9 fy pump wells full, tornado V Reservoir level at El. 695.0 1.67 x Normal per ACI 1.23 1.10 (25-year flood), operating fs = 0.9 fy loads, earthfill, both pump wells full, safe shutdown earthquake (SSE)
WBN TABLE 3.8.4-8 (SHEET 2 of 2) INTAKE PUMPING STATION LOADING CASES, ALLOWABLE STRESSES, FACTORS OF SAFETY, AND MATERIAL PROPERTIES (Cont'd) WSD Factors of Safety Case Description Allowable Stresses Overturning Sliding Va Reservoir level at El. 732.8 x Normal per ACI 1.11 1.10 operating loads earthfill, both fs = 0.9 Fy pump wells full, SSE VI Construction condition-dead 1.33 x Normal per 14.5 4.0 load of structure, no equipment, ACI earthfill, no ground water VII Reservoir level at El. 742.6 1.67 x Normal per 1.512 4.345 (739.2 PMF + 3.4 ft wave run ACI up) operating loads, both pump fs = 0.9 fy wells full. (The electrical equipment room begins to flood when the reservoir level exceeds El. 728) ACI = Allowable stresses per ACI 318-71 Edition (working stress design). Material Properties Concrete: f'c = 3000 psi w = 145 pcf Reinforcing Steel: fy = 60,000 psi (ASTM A615, grade 60)
WBN TABLE 3.8.4-9 CONCRETE RETAINING WALLS LOADING CASES ALLOWABLE STRESSES, FACTORS OF SAFETY, AND MATERIAL PROPERTIES Case Description-Reservoir WSD Allowable Stresses I Normal Operating condition - level at El. 675.0 Normal per ACI 318-71 IA Same as (I) + Operating Basis fc = 0.45 fc fs = .5 fy IB Same as (I) + Safe Shutdown Earthquake fc = .75 f'c fs = .67 fy II Construction condition - earth pressure, 200 psf fc = .5 f'c surcharge fs = .5 fy Material Properties Concrete: f'c = 3000 psi w = 145 pcf Reinforcing Steel: fy = 60,000 psi (ASTM A615, grade 60)
WBN TABLE 3.8.4-10 SHEET PILE RETAINING WALL DESIGN LOADINGS, ALLOWABLE STRESSES, MATERIAL PROPERTIES Allowable WSD Allowable Allowable Stresses Stresses Case Description Stresses ASTM A36 ASTM A328 I Earth pressure Normal per 0.8* (AISC 18,000 psi plus 200 psf ACI 318-71 allowable) surcharge II Same as I plus Normal per ACI 0.8* (AISC 18,000 psi Operating Basis 318-71 allowable) Earthquake fs = 0.5 fy III Same as I plus 1.67 x Normal 0.8* x Fy 28,000 psi Safe Shutdown per ACI 318-71 Earthquake fs = 0.9 fy
- Reduced allowable stresses are used to provide corrosion allowance.
Material Properties ASTM A36 Steel: Fy = 36,000 psi
WBN TABLE 3.8.4-11 (SHEET 1 of 2) TRAVELING WATER SCREENS (INTAKE PUMPING STATION) LOAD COMBINATIONS AND ALLOWABLE STRESSES Structural Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression(2) Shear I Dead 0.56 Fy 0.38 Fy Live with water at El. 683.0 and 2' 6" head loss Impact from live load II For headframe only 0.56 Fy 0.38 Fy Dead Live with water at El. 683.0 and 2' 6" head loss Impact from live load Snow and ice III Dead 0.56 Fy 0.38 Fy Live with water at El. 713.0 and 2' 6" head loss OBE(1) IV Dead 0.9 Fy 0.6 Fy Live with water at El. 695.0 and 2' 6" head loss SSE(1) V Dead 0.9 Fy 0.6 Fy Live with water at El. 741.7 and 5' 0" head loss Impact VI Dead 0.9 Fy 0.6 Fy Stall at 300% capacity
WBN TABLE 3.8.4-11 (SHEET 2 of 2) TRAVELING WATER SCREENS (INTAKE PUMPING STATION) LOAD COMBINATIONS AND ALLOWABLE STRESSES (Cont'd) Other Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression(2) Shear I Dead Ult 2 x Ult Live with water at El. 683.0 and 5 15 2' 6" head loss Impact from live load II For headframe only Ult 2 x Ult Dead 5 15 Live with water at El. 683.0 and 2' 6" head loss Impact from live load Snow and ice III Dead Ult 2 x Ult Live with water at El. 713.0 and 2' 5 15 6" head loss OBE(1) IV Dead 0.9 Fy 2/3 x 0.9 Fy Live with water at El. 695.0 and 2' 6" head loss SSE(1) V Dead 0.9 Fy 2/3 x 0.9 Fy Live with water at El. 741.7 and 5' 0" head loss Impact VI Dead 0.9 Fy 2/3 x 0.9 Fy Stall at 300% capacity Notes: (1) Acts in one horizontal direction only at any given time and acts in vertical and horizontal directions simultaneously. (2) The value given for allowable compression stress is the maximum value permitted when buckling does not control. The critical buckling stress. Fcr, shall be used in place of Fy when buckling controls. Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or E 2 Kl Fcr = when > Cc Kl 2 r 2 r
WBN TABLE 3.8.4-12 (SHEET 1 of 2) DIESEL GENERATOR BUILDING LOADS, LOADING COMBINATIONS, ALLOWABLE STRESSES, AND MATERIAL PROPERTIES Loads D = Dead load of structure including the weight of the diesel generators L = Live load - 200 psf or equipment load in mechanical areas
- 300 psf in electrical areas
- 20 psf on roof Lc = Construction live load (50 psf on roof)
E = Operational basis earthquake (OBE) E' = Safe shutdown earthquake (SSE) WT = Tornado-generated missiles(1)
WBN TABLE 3.8.4-12 (SHEET 2 of 2) DIESEL GENERATOR BUILDING LOADS, LOADING COMBINATIONS, ALLOWABLE STRESSES, AND MATERIAL PROPERTIES (Cont'd) Load Combinations Case Description(3) Allowable Stresses I D+L Normal stresses(2) II D+Lc Normal stresses(2) + 33% III D+L+E fc = 0.45 f'c IV D+L+E' fs = 0.50 fy fc = 0.75 f'c fs = 0.90 fy V D+L+Wt fc = 0.75 f'c fs = 0.90 fy Material Properties Concrete: f'c = 3000 psi and (4000 psi for vent hoods) w = 145 pcf Reinforcing Steel: fy = 60,000 psi (ASTM A615, grade 60) Notes: (1) The exterior walls and roof are designed to resist missile spectrum A of Table 3.5-7. The precast concrete bulkheads placed in front of the equipment doors are designed to withstand tornado-generated missiles of missile spectrum B in Table 3.5-8 (see discussion in Section 3.8.4.1.2). (2) Normal stresses are given for working stress design in ACI Code 318-63 or ACI Code 318-71 (See Section 3.8.4.2.2). (3) Both conditions of L, having its full value or being completely absent, are checked.
WBN TABLE 3.8.4-13 (SHEET 1 of 2) DIESEL GENERATOR BUILDING DOORS AND BULKHEADS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Structural Parts Allowable Stresses(psi) No. Load Combinations Tension and Compression(4) Shear Door Open or Closed I Dead load 0.6 Fy 0.4 Fy Door Closed I Dead Load plus OBE 0.6 Fy 0.4 Fy III Dead load plus SSE 0.9 Fy 0.6 Fy Mechanical Parts Allowable Stresses (psi) No. Load Combinations Tension and Compression(4) Shear Door Open or Closed I Dead load Ult 2 x Ult 5 15 Door Closed II Dead load plus OBE 0.6 Fy 0.4 Fy III Dead load plus SSE 0.9 Fy 0.6 Fy Concrete Bulkheads Allowable Stresses No. Load Combinations Concrete Reinforcing Steel I Dead Load 1.0 ACI 318 1.0 ACI 318 II Dead Load plus Wind(2) or OBE 1.0 ACI 318 0.5 Fy III Dead Load plus SSE 1.67 ACI 318 0.9 Fy (3) (3) IV Dead Load plus Tornado(2)
WBN TABLE 3.8.4-13 (SHEET 2 of 2) DIESEL GENERATOR BUILDING DOORS AND BULKHEADS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Notes: (1) Acts in one horizontal direction only at any given time and acts in vertical and horizontal directions simultaneously. (2) The steel doors and steel bulkheads are protected from wind, snow, ice, rain, tornado, and wind and tornado missiles by precast concrete bulkheads as discussed in Section 3.8.4.1.2. (3) The structure may be allowed to yield for load combination IV when considering impactive loads from missiles. (4) The value indicated for the allowable compression stresses is the maximum value permitted when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls. Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or 2 E Kl 2 Fcr = 2 when > Cc Kl r r
TABLE 3.8-4-14 DELETED
WBN TABLE 3.8.4-15 PRIMARY AND REFUELING WATER PIPE TUNNELS LOADS, LOAD COMBINATIONS, ALLOWABLE STRESSES, AND MATERIAL PROPERTIES Loads D = Dead load of structure plus any permanent load contributing stress, such as vertical soil pressure, hydrostatic pressure from ground water Elevation 710, walls; Elevation 726, uplift on slab. L = Surcharge loading from trucks and other equipment operating above tunnel. D' = Hydrostatic pressure from reservoir Elevation 729, maximum level buildings remain unflooded. Ro = Temperature effects on pipe restraints inside tunnel. E = Operating basis earthquake. E' = Safe shutdown earthquake. Yr = Pipe restraint load due to main steam pipe rupture. Wt = Tornado missile striking earth above top slab of tunnel. Case Load Combination Allowable Stresses I D+L (Construction Condition) fc = .45 f'c (ACI 318-71) fs = .5 fy II D+L+Ro+E fc = .45 f'c fs = .5 fy III D'+L+Ro fc = .75 f'c fs = .9 fy IV D+L+Ro+E' fc = .75 f'c fs = .9 fy V D+L+Ro+E'+Yr fc = .75 f'c fs = .9 fy VI D+L+Ro+W t fc = .75 f'c fs = .9 fy Material Properties Concrete f'c = 3000 and 4000 psi w = 145 pcf Reinforcing Steel: fy = 60,000 psi (ASTM A615, grade 60)
WBN TABLE 3.8.4-16 (SHEET 1 of 2) CLASS 1E ELECTRIC SYSTEMS STRUCTURES LOADS, LOAD COMBINATIONS, ALLOWABLE STRESSES, AND MATERIAL PROPERTIES STRUCTURES - Manholes Design Cases Allowable Stresses I. SEISMIC OPERATING
- a. Dry earthfill plus 1/2 safe shutdown earthquake fc = 0.45f_c (1/2 SSE) fs = 0.50fy
- b. Earthfill with ground water at elevation 726.0 or fc = 0.45 f_c finished grade, whichever is lower, plus 1/2 SSE fs = 0.50 fy II. FULL SEISMIC
- a. Dry earthfill plus SSE fc = 0.75 f_c fs = 0.90 fy
- b. Earthfill with ground water at elevation 726.0 or fc = 0.75 f_c finished grade, whichever is lower, plus SSE fs = 0.90 fy III. NORMAL OPERATING Earthfill with ground water at elevation 726.0 or finished fc = 0.45 f_c grade, whichever is lower, plus 200 psf surcharge (or fs = 0.40 fy concentrated surcharge where applicable)
IV. TEST Dry earthfill, one compartment of manhole filled with fc = 0.50 f_c water, water surface elevation 743.5 fs = 0.60 fy V. FLOOD Earthfill plus probable maximum flood. No water inside fc = 0.75 f_c structure fs = 0.90 fy VI. TORNADO LOADING D+L+WT(Vertical Missile) See UFSAR Section 3.5 D+L+WT(Differential Pressure) fc = 0.75fc fs = 0.9fy
WBN TABLE 3.8.4-16 (SHEET 2 of 2) CLASS 1E ELECTRIC SYSTEMS STRUCTURES LOADS, LOAD COMBINATIONS, ALLOWABLE STRESSES, AND MATERIAL PROPERTIES (cont'd) STRUCTURES - Duct Banks Design Cases Required Strength I. SEISMIC OPERATING 1/2 SSE (E) U = 1.4D + 1.7L + 1.9E II. FULL SEISMIC SSE (E_) U = D + L + E_ III. TORNADO GENERATED MISSILES U = D + L + WT IV. SURCHARGE LOAD L (CRANE OR TRAIN) U = 1.7L + 1.4D (D = Dead Load) Material Properties Concrete: F_c= 3000 psi w = 145 pcf Reinforcing Steel: Fy = 60 ksi (ASTM, A615, Grade 60)
WBN TABLE 3.8.4-17 NORTH STEAM VALVE ROOM LOADING COMBINATIONS AND ALLOWABLE STRESSES WSD Allowable USD Load Factors Load Combinations Stresses Case I = D+L fc =.45 f'c 1.4D+1.7L fs =.40 fy Case II = D+L+E fc =.45 f' 1.4D+1.7L+1.9E fs =.50 fy Case III = D+L+E *fc =.75 f'c 1.0(D+L+E) fs =.90 fy Case IV = D+L+W t *fc =.75 f'c 1.0(D+L+W t) fs =.90 fy Case V = D+L+Pa *fc =.75 f'c 1.0D+1.0L+1.5 Pa fs =.90 fy Case VI = D+L+Pa+Yr+Yj+Ym+E *fc =.75 f'c 1.0D+1.0L+1.25 Pa + fs =.90 fy 1.0(Yr+Yj+Ym) +1.25E Case VII = D+L+Pa+Yr+Yj+Ym+E *fc =.75 f'c 1.0(D+L+Pa+Yr+ Yj+Ym+E) fs =.90 fy
- Concrete stresses other than flexure = 1.67 x normal The following terms are used in the load combination equations:
D - Dead load of structure and any permanent equipment loads or hydrostatic loads L - Live loads, including any moveable equipment loads such as soil pressure E - Operational Basis Earthquake E' - Safe Shutdown Earthquake Wt - Tornado, including wind pressure with missiles Pa - Pressure from postulated main steam pipe break Yr - Pipe anchor force due to postulated pipe break Yj - Jet force due to postulated pipe break Ym - Missile impact force due to postulated pipe break
WBN TABLE 3.8.4-18 (SHEET 1 of 2) NORTH STEAM VALVE ROOM STRUCTURAL STEEL LOADING COMBINATIONS AND ALLOWABLE STRESSES Loading Tension Shear on Gross Compression on Combinations Net Section Section Gross Section Bending Concrete Bearing Case I 0.60 FY 0.40 FY See Note 1 0.66 FY to 0.25 f'c DL+LL+OBE 0.60 FY Case II 0.90 FY 0.9 Fy See Note 2 0.595 f'c 0.90 FY 3 See Note 3 DL+LL+SSE See Note 3 Note 1 - Varies with slenderness ratio; see AISC "Manual of Steel Constructions," 7th Edition, Table 1-36, page 5-84. Note 2 - Varies with slenderness ratio: ( KL / r ) 2 (A) Fa = 0.9 Fy 1 2Cc2 Main and secondary members where KL/r Cc: 0.9 2 E (B) Fa = Main members where Cc < KL/r < 200: (KL / r ) 2 Fa [ byFormula ( A)or ( B)] Fas = Secondary members where 120< KL/r 200: . L / 200r 16 where: 2 2 E Cc = F Fy E = Modulus elasticity of steel Fa = Axial compressive stress permitted in the absence of bending moment Fas =-Axial compressive stress permitted in the absence of bending moment, for bracing and other secondary members.
WBN TABLE 3.8.4-18 (SHEET 2 of 2) NORTH STEAM VALVE ROOM STRUCTURAL STEELLOADING COMBINATIONS AND ALLOWABLE STRESSES FY = Specified minimum yield stress of material (kips per square inch) f'c = Compressive strength of concrete (kips per square inch) K = Effective length factor L = Actual unbraced length (inches) r = Governing radius of gyration (inches) Note 3 - When the supporting surface is wider on all sides than the loaded area, the permissible bearing stress on the loaded area may be multiplied by A 2 / A1 1 but not more than 2. Where: A1 = Loaded area (square inches) A2 = Maximum area of the portion of the supporting surface that is geometrically similar to and concentric with the loaded area (square inches).
WBN TABLE 3.8.4-19 ERCW STRUCTURES LOADS, LOAD COMBINATIONS, ALLOWABLE STRESSES, AND MATERIAL PROPERTIES LOADS D = Dead Load L = Live Loads (loads which vary in intensity and occurrence) E = Operating Basis earthquake (one-half safe shutdown earthquake) E' = Safe Shutdown Earthquake Wt = Tornado Loading (Wind and Missiles and pressure differential as applicable) W = Loads Generated by the Design Wind for the Plant Lc = Construction Live Load
= Structure - Slab and Beams Supporting ERCW Pipes LOAD COMBINATIONS ASTM A36 Design Cases Allowable Stresses Structure - Slabs Supporting ERCW Pipes I u = 1.4 D + 1.7 L 0.6 yield II u = 1.4 D + 1.7 L + 1.9 E 0.6 yield III u = D+L+E' 0.9 yield Structure - ERCW Standpipe Structure I u = 1.4 D + 1.7 L + 1.9 E II u = D+L+E' III u = D+W t IV u = 1.4D + 1.7L + 1.7W Structure - ERCW Discharge Overflow Structure I u = 1.4 D + 1.7 L II u = 1.4 D + 1.7 L + 1.9 E III u = D+L+E' IV u = D+W t VA u = 1.4 D + 1.7 L + 1.7 W VB u = 1.2 D + 1.7 W VI u = 1.4 D + 1.4 Lc Structure - ERCW Valve Covers I U = 1.4 D + 1.7 L II U = 1.4 D + 1.7 L + 1.9 E III U = D+L+E' IV U = D+W t MATERIAL PROPERTIES Concrete: f'c = 3000 or 4000 psi w = 145 pcf Reinforcing Steel: fy = 60 ksi (ASTM A615, Grade 60)
WBN TABLE 3.8.4-20 REFUELING WATER STORAGE TANK FOUNDATION LOADS, LOAD COMBINATIONS. AND MATERIAL PROPERTIES LOADS D = Dead Load L = Live Load Including Soil Pressure E = Operating Basis Earthquake W = Design Wind E' = Safe Shutdown Earthquake Wt = Tornado Loading (Wind and Missile) Yj = Jet Impingement Associated with High-Energy Pipe Break Ym = Missile Impact Generated by High-Energy Pipe Break LOAD COMBINATIONS Design Cases 1 U = 1.4 D + 1.7 L 2 U = 1.4 D + 1.7 L + 1.9 E 3 U = 1.4 D + 1.7 L + 1.7 W 4 U = 1.2 D + 1.9 E 5 U = 1.2 D + 1.7 W 6 U = D+L+E' 7 U = D+L+Wt 8 U = D+L 9 U = D+L+Yj+Ym+1.25 E 10 U = D+L+Yj+Ym+E' MATERIAL PROPERTIES Concrete: f'c = 3000 psi w = 145 pcf Reinforcing Steel: fy = 60 ksi (ASTM A615, Grade 60)
WBN TABLE 3.8.4-21 SPENT FUEL POOL GATES LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Allowable Stresses lb/in2 Loading Conditions No. Load Combinations(1) Bending Shear 1 D+L 0.6 Fy 0.4 Fy 2 D+L+OBE 0.6 Fy 0.4 Fy 3 D+L+W 0.6 Fy 0.4 Fy 4 D+L+To+Ro+SSE 0.9 Fy 0.6 Fy 5 D+L+To+Ro+W t 0.9 Fy 0.6 Fy 6 D+L+Ta+Ra+Pa 0.9 Fy 0.6 Fy Notes: (1) To, Ro, Ta, Ra, Pa = 0
WBN TABLE 3.8.4-22 DELETED
WBN TABLE 3.8.4-23 (Sheet 1 of 2) WATERTIGHT EQUIPMENT HATCH COVERS LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES Allowable Stresses (psi) No. Load Combination Tension Compression* Shear Hatch Closed 2 I D + 200 lb/ft live load 0.6Fy 0.6Fy 0.4Fy II D + L1 0.9Fy 0.9Fy 0.6Fy III D + L2 0.9Fy 0.9Fy 0.6Fy IV D + L1 + OBE 0.6Fy 0.6Fy 0.4Fy V D + L1 + SSE 0.9Fy 0.9Fy 0.6Fy Where: D - Dead Loads or their related internal moments and forces including permanent equipment L1 - Live Load due to flood to El 711.0 L2 - Live Load due to pressure of 3 psi from below OBE - Loads due to the operating basis earthquake SSE - Loads due to the safe shutdown earthquake
- The value indicated for the allowable compression stresses is the maximum value permitted when buckling does not control. The critical buckling stress, Fcr, shall be used in place of Fy when buckling controls.
Kl 2 r Kl Fcr = FY 1 - when < Cc 1 2 Cc 2 r or 2 E Kl Fcr = 2 when > Cc 2 Kl r r
WBN TABLE 3.8.4-23 (Sheet 2 of 2) WATERTIGHT EQUIPMENT HATCH COVERS Serial Designation of the Material Specifications of the ASTM Structural Steel A36 Pipe A53 or A103 Grade B Headed Concrete Anchors 1/2" diam x 5-3/16", A108 Steel Screws A193, Grade B Seals Natural or synthetic rubber or combination of natural and synthetic rubber (This is not an ASTM designation)
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REACTOR, AUX & CONTROL BLDGS CONCRETE GENERAL OUTLINE FEATURES TVA DWG NO. 41N700-1 RB FIGURE 3.8.4-1
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-2
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-3
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-4
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-5
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-6
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\ 3, Z. COMPLETION OF DESIGN FOR COVERS AND OPERATING EQUIPMENr BY CONTRACTOR TO PROVIDE OPERIUOR AS OUTLINED IN 1VA SPECIFICATION 1437.
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- 9. THE SHACKLES BETWEEN THE WIRE ROPES AND THE 0-RINGS
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~--+-+-~ ArEL729'-0" WATTS BAR FINAL SAFETY ANALYSIS REPORT lJ'-6 , .. AUXILIARY BUILDING COMPANION DRAWINGS,' .f4N:S26, UNITS 1 & 2 ELEVATION 8-B -:SZ T, a-:SZ8 RAILWAY ACCESS HATCH ARRANGEMENT & DETAILS SH DUAi III ASSURANCE t HATCH COVERS AND FRAME ARE CLASS I SE I 5141 C EQU I PM[NT. ALL MATEUA.l FOR CLASS I EQUIPMENT REOUIRES C0'1PLETE QUALITY ASSURANCE t>OCUp,t[NTATION. FOR WELDING BY TVA FIELD, WELOIJf(; AND INSPECTION OF NOT TD SCALE EXCEPT AS NOTED TVA DWG NO. 44N325 RF WELDS OIJNE Ill ACCORDAflCE WITH GENERAL CONSTRUCTION SPECIFICATION G*29.
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WATTS BAR Orum Pif~h D1~,,,rrtrt (Mln.) IS* FINAL SAFETY iii/II- Nope l)l~(Stt Spcif'i~,1tio,,J t* ANALYSIS REPORT e1.e VAi/ViV &.NV t.L.t. VAi IUIV AUXILIARY BUILDING Doo,, .J(HH (RM1ing) ll-lO fp,n UN ITS 1 & 2 RAILWAY ACCESS DOOR HOIST MACHINERY ENCLOSURE DETAILS TVA DWG NO. 44N318 RF FIGURE 3.8.4-15
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), FIElD REQUIREMENTS:
A, MATERIALS REQUIRE ONL'f' OOCUJilllENUTION: HEADED CONCRETE
,M:!CHORS, ASH~ A !09; PIPE FITTINGS .ut:D PIPE, ASTN A 106, GR B.
ii, WELDING o, CONC!ii:Eft ANCMOK Slues AND IPll'IN&, AND IN$jl[Cft0N AHO TE ST! NG Of' THESE ~LOS ! H .1,CCOR:D,1,fri!CE W! TH GEN CONST SPEC G-29C. WELDING FOR 251 OF CONCRETE ANCHOR STUDS TESTED ON A Fi.1Uii'.lVP'i BASiS PER G *21C, WHW FOR OTMC" .ili'iCMOR STUDS TO BE VISUALLY INSPECTED.
- 4. ALL MATERIALS BY TVA FIELD & FIELD WORK SHALL BE IN ACCORDANCE WITH
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WATTS BAR FINAL SAFETY ANALYSIS REPORT DETAIL D DET,~/L £ AUXILIARY BLDG Full Sea;. UNITS 1 &. 2 MANWAYS RHR SUMP VALVE ROOM TVA DWG NO. 44N355 RJ FIGURE 3.8.4-16
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-17
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-18
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-19
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-20
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C, w z ,_, ...--- - EACH S11X. PANELS TO ANO ARE APPi/OX 4' x 7' HIGH EACH OPENED. < 8£ INTERCHANGA8'.E 1-z CARO REAOeR Rl:0/JIRED ,_, ON OOORS A/19,All,I, Al52, < Al5S, AIS8, A/59.
- IE C,
DETAIL A DETAIL1\
/
\ !NA CTIVE LEAF u DOOR A//J ONI..Y (44N3'!)
SU OETAIL C (44NJ61)11 DOOR All~ !NTEll!OCJ<EO WITH A/12 FOR lt?WOW. IXXYI A/15 DOOR Al2J INTERLOCl<f:0 WITHAl22 ONLY DOOR AIJZ INTEIILOCl<eO WITHAIJ,J DOOR AIS2 INTERLOCKED WITHAIS/
...---HINGE' (iyp\ , I NOTE:
THE INTERLOCK LATCH CHANISM FOR DOORS A77, A154, A156, A17J, A191, .t: A214- AS SHOWN DOORS A65 AND A7 8 BE SO TAGGED AND WILL BE INSTA TVA FIELD. A132,
!£CHAN ISM FOR OT ARE TO BE INSTALLED BY DOOR
- 7. SEE 461501-SERIES FOR ABSCE BOUNDARIES.
~,.~,Nl.,=--r~, " ~~-T~
-rr=- --- -- - -:\ f"=--:
13lr III_
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'4,,,/Gfl!NO SMOOTH
- 1. DOOR ARE CLASS 1 SEISMC EQUIPMENT.
FIELD
- 1. MATEfllAL FURNISHED BY FIELD IS NOT STRESS CA."lflYlNG MATERIAL AND THEREFORE REQUIRES NO QUALITY ASSURANCE DOCUt.t::NTATION,
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- 2. SEE CONDUIT AND GROUNDING DFIAWWGS FOR CARD READER LOCATION
.UJn\AIIRIMf:.,
- 3. FOR SEAL MOOFICATION BY TVA FIELD, SEE TVA DWGS 44N409*1, j 'rop OFl"xr- / EACH ENO/ I A.;.~~::~~ C!s:'..1.1 1/2*~ 3/4"U*v Rs:' ~nDAs:'U~ l"'l'\I..U)Rl=C!~n tI DOOR 5 TOP '-roP OF ('x ,*
C6xtJ.ol/ GASK£T AHO MAY BE MOUNTED ON MEAD.JAMBS.DOOR BOTTOM DOOR STOP OOOR SEAL LOCI/ING ll j* (NO OR ASTRAGAL & SOLI> NEOPRENE SUCH AS ZERO. NO, 326A
.t.TTJ.t"l-lrn .lT IVY'jl RnTTr&A .l~ ~01111:11:n Tn U.llfilT,'-1.1 CIRl=.ltlitlSII= RnlllJl"llll!ES f*xkff SHIMS 8H!NO I I '-Vl...nlfVV /CJ
/Y I * ,
'GUSSET Ii 'S /"-"X-'-" X 0'*ii?"
DETAIL A a 4 I I OOORLEAFNOTSHQWN SECTION A-A 11 OV£RALL FRAM£ W/DTH!,t," B-B LOCl(JNG ii NOT SHOWN HTS UFSAR AMENDMENT DETAIL B EDGE or f?(MO 'v'AOLE 2 NTS WATTS BAR FINAL SAFETY ANALYSIS REPORT PROV/DC f'"Vl,.,-,.C.I WATERPROOF r,, .C.,...I, ,-, ,..,t.'f,..C. NOT TO SCALE EXCEPT AS NOTED AUXILIARY BUILDING TO KEEP CONCP£TT AWAY FROM BOLTS UN ITS 1 & 2 RAILWAY ACCESS HATCH ARRANGEMENT & DETAILS-SHEET TVA DWG NO. 44N36O RL FIGURE 3.8.4-21
(!) z 3: a::: Cl Cl w z ~*--
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- I: ~Hl!:_G-=J Cl PIN 8-8 & 81-81 T u<
I 81-81 AS NOTED
~
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lJ TUBE4X2X J* TUBE3X2X ft" A77 A132 A214
*/-3/8" ii J// _ ','.i ~ i 2. SEE 46W501-SERI£S FOR ABSCE BOUNDARIES .
...c_",v,,, r, ,,a __ ,,.
*1 FRAME TYPE I ....1.1 I l"
_u_-_-_-_-_~_::1_____..:_~_1._.___..!
~"'rJ I !!
OCKJRS A12J ct ,41,12 ONLY $EE St:HEt:/1.Jl.E I/E/;IOVA4U PANELS NOT SHOWN HTS De.1AiL C SC.ll.LE: 1/2n = 1'-0" WiNOOW ... 000,li AtiSOIYLY SCALE r= r-o* 0 WATTS BAR FINAL SAFETY ANALYSIS REPORT POWERHOUSE SCALE, 3" = 1'-0" AUXILIARY BUILDING EXCEPT AS NOTED PRESSURE CONFINING PERSONNEL DOORS ARRANGEMENT & DETAILS-SHEET 2 TVA DWG NO. 44N361 RL FIGURE 3.8.4-22
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(.) l+odde .#'op a Mceas*ry lo ~~*,,,.,. M+t:h lrom belnJ l>',lllcd ou-f or gu,~e~ Outline of door latch cover, for door, A 156 S£CT!ON A-A 8-8 A77 I /b-2
- A214 Nono
------*----- A/32 Non~
E-£ i:. ii)
-L -H' l"'++---"
iT i / j_i~_ _....,..i--)~ ALL Loc.r.S l(EYED 17b Alff TO BE l(fYED ALJl(f IN GROUPS A. SUBJECT ro GRAND MASTER 8. 0
'-SOUO m.iBBEfil 1"
_1___ i i i I i j_ NOTES /FOR DOOR Al62l THICK X 3"X 6" CYLINDER RIM DEADLOCKS ARE NOT REQUIRED FOR DOORS IOll,I;~, IOll~.i;, ,..,,_,..... 1"11._....., I'll...,_,, l"\lf~, ... ,o-., ... , : . 1 , 1 * (TYPl.J l-_ 1.FELO (OCATE DOOR STOP K-K A214. AND A215.
** K ~1 1_ N!E.-'-JI EDGE OF DOOP ,
NTS _,.. DETA.iL C t~...e~E~tfcf~OOriED FOR W.8.'S ONI..Y, TVA wit I furn/ah name of vendor and manufacturer providing key 1A, grand masier key A, muter 8-1, l.* NOTEA. ,,,___, a-,, .,,..-ru 11-nas lJ, ar,u . , , _ , lif'llnV ,,,....,. SHEET -f."TH/1.. (AST/1,f A570) SEE NOTE A (BY TVA FIELD)"-\ SEE
/r-;v-< To v:JOTH r l " A_C: _f:IJl"tYIAJ Tl'ti FIT AGAJNST SEAL BAR.
TVD Rr'ITU C'&lnr.
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,,tr . . .\.!_.. ! :..!.." 1 -". UIJTf:RJJJ, .C: .aNn IN~/ 1)/IJ(; FW TVA FIELD IN ACCORDANCE
-'----+"I+- f ' WffH OUALm' LEVEL ll I~ I : I: I* T L3 x 2 x-!.-OR I ~ :Ii l ru: rruJ.t:TQJlrTlnAJ - ~ r .
N3G-81l1.
- 3. FIELD WEl.iliNG AND WATTS BAR l!
*ii IE : : : . ! J....L ~I EQUAL (ASTM-A36 l \ ' : i : j g IJJ.c:~rTIAJr. ~ w~,.n.~ '"
1h ! I : _,,t~J :1 TYP 1 CUT AS SHOWN,
- I ~ ! I! : U ACCORDANCE WffH GENERAL FINAL SAFETY
\_ See FLOOII ELEV llf:l:.l:;:1;:::.~1,1 1 (BYTVAFIELD)..._/ ~CR" -- U
- CONSTRUC1iON SPE.CiFIC~liON SCneuu~ c441'f.3o, J li' GT I ~'I 1 I REMOVABLEI0G.A. ~:,, II M ::
G-29C. ANALYSIS REPORT 4 rr.r: :i7 1!1,~flttii2f 'j II i"Fi.ANGE
,rp
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i PL(ASTMA57oJcovER.&iti...JI TYPICALFORBOTT0/1,f l!-i-~~ ulruu OOOfl 1:1: i4l ~ 1 ~~ , COVER EXCEPT OPP ____bl I** ~,u--- POWERHOUSE rr.t==fiw'~
- HA.'ID (BY Tl.IA FIELD) 7ii'i'i11'i!r':i!w'r~.fl.::ii-.--_---ii-----'-
AUXILIARY BUILDING
, A4-t-JII. ""itr \.3/4*-2a uNF.ROUNo Ho
... i::.1 111..1 J I SCREW)( t* LG. TYP 7 PLACES.
Anrs"*,___/ 11m u 11 I :r SCALE: 5"*1'-0" PRESSURE CONFINING
. F-F & F1-F1 1- 6" ! I ~_Rlf.! & !A!'~ REQ'D.
, V REQ'D . I!! : :1 EXCEPT AS NOTED PERSONNEL DOORS ARRANGEMENT & DETAILS - SHEET 3 TVA DWG NO. 44N362 RF FIGURE 3.8.4-23
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-24
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-25
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-26
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-27
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-28
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-29
PPIMARf'Ll"H MK44N,?.26*/
'Z0' ScCTION A-A
[)IE5£L GENERATOR SU/LOIN(; r l3 TO l3 C's *N()M l--------~9_**~6'_B=A=CK~TO~lJACK EMBEOOCO S T E E l ~ - - - -
>----*~*--1*
-,*-1:1
- MAX MA,r C!:f.Ajl OPl'N!NIS PPIMAliY LEAi' OPEN CU-AR OPENINIS
- BOTH UAVES OP[N
~r 19 TOP BOTTOM LATCH PINS&
7~ 1 (f/1!Jlfffi° STEEL L-.,;i::::::::=::::::i::::IJ...ll. A
- A::N:~:::.... I
- 1.
- 4 NOMCLEA/f l'-'f"fiMA/"f'-,- Le.If/' ..::,cc. v1Vu,,,,,-,r ,,,..,,.
M.'<4-l.lt/226 */ ,.+f,'<44N~26 *2 WATTS BAR FINAL SAFETY ANALYSIS REPORT DIESEL GENERATOR BUILDING UN ITS 1 & 2
£LVAT/Ol c"'c £/..£VAT/ON £ *F t:.Lt:.WHICHV /' "/' COMPANION o~s' DOORS & BULKHEAD 44N226 & 44N227 ARRANGEMENT TVA DWG NO. 44N225 RD FIGURE 3.8.4-30
t, z 3' a::: Cl r:---~ Cl w z ------ 1-z
- E DETAIL A
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< TYPICAL B-B u SCAll 3"*1'*0" $CAU 3"*1'--()" 4*-1f*o-o ,,-r*
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** 4#" 12 SPCS
- 4"-4'--()" , ,~f ,ti FLANGED BUSHING REQD - SELF LIJBRIUTWG, GRAP/IITE IMPRE '1,ATED 8RoNZE - MK 3 WASHER 16 li'EOD*SUF J.1/Bli'ICATIN{J, G/1APH/T£ IMPR[GN4T£0 SHIM ASnt A569 *H/5 SCAU 3"=1'--()"
SCALE 6"*1'-0" ~li'ON'-£*/lfK~ r ,~,\.~~~~c~,--~(t~ ~ ~~~& '..J. I 1 .t ONMK20NL'I' SC'4L£ ,*. l'-0
- I
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l't'--+--r-1-'- I LC:~- I ~ I 1'-JI I I DOOR lEA,1:=MK/42 A 1:u:r.-,, _ u,1 A<' *un .,,,.,.en i)-i) E-£
<'J.U'WUill SHOWS tX>OR F.l?AM! ONLY
;f REQ'D - Ml{ Z OPP HAND AND NOrED WATTS BAR FINAL SAFETY ANALYSIS REPORT DIESEL GENERATOR BUILDING UNITS 1 &. 2 DOORS & BULKHEAD DETAILS SHEET 1 TVA DWG NO. 44N226 RD FIGURE 3.8.4-31
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HAff PLAN DETAIL A SUTIOIV A*A B-B c-c 0*04,lJ'*lJ' Tr'PICAL TYPICAL SCA£( 3"=1'-o" SCAt.£ 3"*1'--0" SCAJ.£ 3"*1'-0" SCAff 3"=1'-0" D*O AS SHOWN £-£ D'*D' AS NOT(D T'f'PICAL*J LIJ(;S SC.AL£ 3"~1'-0"' SCALI: 6"*i*O*'
~ R l t NUMIU.S TO Ml.YE HEFII "&tNU7 UNLUS DTHE~lst
- i; 0,1110 2 DUAllU ON THIS DIUWIN,. _ -----*- **-- *-*-
~. ! :.. ~n.~r~ ?i.t*J~lt~1.. L~:~n C0R"OS IOfl*IIIU l':iTUr ptU l'tU-11 MK."i. ':"S,&*"o-iURNiuili£:'"Jlw -.No JAW ENDS, VITH 112" D
- !~*!ij,.~-..~0ui!~~"~J.~l!~~~L~E f J~;*l1: ~~"ii6o.
a* r---- .-,., WATTS BAR I ~, r
.!.,ft'I d"oHOl£;
;ol MK,___.,-,,
I FINAL SAFETY ANALYSIS REPORT DIESEL GENERATOR BUILDING I I UNITS 1 &. 2 DOORS & BULKHEAD COMA4NIQN D"'l>S*
. ***--- - . ***--- DETAILS SHEET 2 TVA DWG NO. 44N227 RD FIGURE 3.8.4-32
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Cl (..) II 't ij-"fJ HOLE fN COVER 87-87
~
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II EL 750.0S
.. TYPICAL PRECASTBARRIER STRIP
.32 REQUIRED NTS t FOR3/4" WBAND 11." iJ HOLE IN COVER 16 t FOR3 U.C.
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17 INSTALL SHIMS BETWEEN THE 4 X 4 X ANGLES AND THE PLATE1/2 X 5 N,c<S!>ARV TO PROVIDHD"1ACT. TAL,K 1/2AS j ANCHOR BOLT TYP (SEE A7-A7) KEY PLAN NTS NTS J'~~i.mo*****1 i,,;;-,;; DETAIL A7 0 (EXJSTING) NTS ~ & 3/4"¢ANCHOR (NE"rl) NOTE El= 1.,r- 1 ,n,;i 1.u,;11,;o HnU nc~,;,,xa., ,.......,..,1'.C, ,;,:
?.."'6=......""-r---~:t,SZ//f:? *****
0 I 1);:~)~~ '::t E/;!fililt:i.~flicfa/JR~titt:J#JiifAH!C L.i:..J <:.:1~.u_.11-,u, -* -*****- I NfAl'IUrRI.,/ uneno Al"l"Ur..;Rj IVN HW{)/l'rUl.,//VNO. Nf/N/NfVM :,unr.At,e PREPAFIATION SHOULD BE SSPC SP3" POWER TOOL CLEANING"
' *zs-B I i"s-8' oNi.i .., NOTCH ANGLE LEG TO Z) THE DEPTH OF THE UFTING LUG RECESSED POCKETS MAY BE ONLY ,.nvv11.1c7t~ mH'f 1,,,1.11 CH::fn'CCf"f nC.UUl.,CL/ er rfl.LHWU " " " LIi.in' 1.,Vl'rfY/fYU i)fUl.,UfYC n,v rv>tm i:>CRI.
~Ii/CT, rnAJNJJT~ ,1.un JJA,11:.11:*. 13-6518 OR EQUAL PRIOR TO COATING WITH AMERTHANE 487.
ZB-BONLY CAUTION SHALL BE USED - NOT TO lfiTERFERE WITH REQUIRED
<..,U:AtfRN<..,e rvn urnr,G uev11.,e, DETAIL Y...7 (3 REQ'D) 3) 1111:'TJUf'.! 111,:. ar:rl:'c.-c.-J:'n t:>l'V'Jtrrc.- - APPLY AMERTHANE 4!7 nrTJIJJ tJl?(t GJ:'tvn)(:>11-11) ( ELASTOMERfC POLYURETHANE) TO ALL OF THE RECESSED SURFACES
't CURB, REBAR, BOLT & SSD
,. I ,. I,. ;:;s n,uc.-. 3/4*,, u,......
AND TIE-IN TWO (Z) INCHES ON LIFTING LUGS AND EXTERIOR CONCRETE SURFACES iN ACCORDANCE Wi TH THE MANUFACTURERS APPLICATION INSTRUCTIONS. MINIMUM SURFACE PREPARATION II SHOULD BE IN ACCORDANCE WITH ASTM 4258 "STANDARD PRACTICE II J\--r _ L7-, rvn .::;<.mr.<11.,e 1.,u:.<1rwtrwu 1.,vtv1.,ne, e run 1.,u.<11 tNUO," II r* 1 tK7 _,-CURB, -l(n , MIN.3/4- CLEARANCE FOR REINFORCING STEEL AT RECESSED POCKETS. Pt.AN
; - - 7 ;-- --;-.;n- - -:J,. NDTES, WATTS BAR
~ - - _Jr 1...- ~" j.) * .Ji '-.~~~-".!:.~~-~ .~: !.."':'.~*-~~~~"t** FINAL SAFETY NOT TO SCALE H7*H7 EXCEPT AS NOTED D£iAiL G7 4 REQ'D, SH3 NTS I
.........., I/
l'-S,E.££1f! Ll Z-CONCRETE CLASS SHALL BE 300.376A 3-REINFORCING STEEL SHALL BE GR 60. COMPANION ()IW'{JN/J.s; ANALYSIS REPORT I 3 I \ ,rr / 4-~f1'1~J}A~~~ll.~ tEE AFICHfrECTRUAL /ON3?0*1 rHRIJ -6 z*-o* I
. 3 '- 4 a" (REI" JI 6-APPLY BONDING AGEfiT To PREPARED suRFAcEs
~AULK._) t:iR~un,~n (REF j I I IN ACCORDANCE WfTH G-34 R4, SECT 4.6. YARD 1/2'.-+tt---,..,,.--,oiifli'iiUii a- DIESEL GENERATOR BUILDING rruum Qi
~I~
- &.fOVV OV/-4MU
);==!fuf-1 Fr:t;I 1----1, CONCRETE CURB_, ~
~BLOC/OVALL
_1_ _ _ _ 1 "t FLOORS&. WALLS OUTLINE-SHEET 7 TVA DWG NO. 10N320-7 RF FIGURE 3.8.4-33
FIGURE 3.8.4-34 DELETED
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-35
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-36
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-36 UNIT 2
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-36A
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-36A UNIT 2
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NOTE;,: If ORY PACKIN& OPTION IS SELECTEO er f!l,O, B4*B4 C4-C4 TNl-5 ANG/.£ REOJ.JIRES A f"/[ll) W£1.0. NOT£$, I. FOR NOTES II RE:FEIIE:NCE: 01/AWINGS SE:£ ~IN363*/. t SELF DI/LL/NG ANCHOR. PH.it/PS 2. ALL MATERIAL AND FABRICATION er TVA FIELlJ. TUNNEL *e* OPP HAND & NOTED VER:r -SU~ORTS -ONLY NOTE F, WATTS BAR FINAL SAFETY ANALYSIS REPORT NOTE. G-4: (UNIT 1 ON!..Y) AUXILIARY BUILDING
,. ('nU.c' Wt:/ TLC:: .C::Ul"IWJJ nJJ TU/_(' nRJJIA/JAJr.:. UNIT 1 (SERIES) WERE FOUND TO BE DEFICJENT BY THE WELD EVALUATION PROJECT, 1-14*114 CONCRETE EVALUMED !fl WATTS BAR ENGINEERING PROJECT AND FOUND TO BE SUITABL£ FOR MJrE C: PIPE TUNNELS & TANK FDNS STRUCTURAL STEEL SUPPORTS TVA DWG NO. 41N363-4 RD FIGURE 3.8.4-368
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C, (.) B4*84 C4-C4 UNIT 2 ONLY: MINIM/JM ACTUAL STRENGTH Of CONCRETE 3500 PSI 2-41NJ6J-1.
*****VZZ/251 DETAIL H OPP HAND V/29/21 (14-(}4 WATTS BAR FINAL SAFETY FOR HATCHED AREAS SEE ANALYSIS REPORT UNIT 1 AC DRAWING 41N363-4 PART-r'L~JfM H4*H4 ALT£.RNAT£ OET.4/1 C4 AUXILIARY BUILDING TIINNliZ 11* /Jf'f' NANO & ICTBJ lCT.!C: FOR wcATI/JNSEc PART*PLANitN&YJ UNIT 2 SEe NO'TE G-4 l'lliZD AMJ SNtJP /1/ElOS A!rE 4tSOl MAN AS5IJNEO ct!K$T.$c"<m"!,t"E.l'JRQ ,IUY,IQF/f CONCRETE Tl/IS SEQIIE/11:E AS ~Sll?EQ. PIPE TUNNELS & TANK FONS STRUCTURAL STEEL SUPPORTS TVA DWG NO. 2-41N363-4 RO FIGURE 3.8.4-36B(U2)
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- i MINIMUM ACTUAL STRENGTH Of CONCRETE 2800 PSI.
Cl tJ WATTS BAR FINAL SAFETY ANALYSIS REPORT AUXILIARY BUILDING UNITS 1 &. 2 CONCRETE PIPE TUNNELS & TANK FDNS OUTLINE TVA DWG NO. 41N363-5 RF FIGURE 3.8.4-36C
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-37
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83 83 I I L I I -':.,,.1 6'!0- 4' .J NOTF (UNIT 2 ONI X)- MINIMUM ACTIJAL STRENGTH OF CONCRETE IS 4600 PSI IN CLOUDED AREAS. WATTS BAR I FINAL SAFETY
. I .
'-+-EL T.26: o+----._ I ANALYSIS REPORT I i '1 I I ! AUXILIARY BUILDING I ! UNIT 2 EXP JT DETAIL I
NOTL FOR HATCHED AREAS SEE UNIT I DWG 4-1N397-2. CONCRETE L.4J c3.,.J NORTH STEAM VALVE ROOM OUTLINE cZZZZZZZZZ> TVA DWG NO. 2-41N397-2 RO COt.lPANION DRAWINGS: 2-+1 N397-1
- 2-+ 1 N397-4. 2-+ 1 N397-3 FIGURE 3.8.4-47(U2)
(.) z
- a 0
- ::
0 0 w z ...... 2*-s* I I , II ! I r l:9,.' I
~
i i b<... I 1-z EL 781,0 en 1s1.o l!L 181.0---...
- Ii I
I I
! i
! _,c-D. 7 77.11 l!L 777.IZ----.._ /
//
0 I I < ,c,Or'TIONAL CONST J Of'TIONIII CONST JT u Et. 775,5 Ei.775,S l!L 775 .5-" I
-- DETAIL 83 I I f;!!!!J I I I cCONST JT EL. 768. 0 I
r----7
,.-EJ.770,25 n.no~r I
..J I
e,CONSTJT EL 7EB.0 EL 170. 25--. CON$TJT n1&e.P I L--------- I I I i i ~-OPT VENT CONST JT ',ll15J,0 TO EL l~I.O - I I I
! ! I I I 7'-tr' L
f-5:' tiROOYE 1/J' I * ..tJ -1..L I I *I , . i*li-CONST .;r,-k_j(~ AS RcQD I "- .,.j
,I. . . . . . r1if}
~1,1/.*i.! I I 'l~ I Tr~l I I
~-1 !* h
- 1 '-CONST JT, AS RE:00
,-VH LVw4IIVl't Xt; o,*o.s 1.2' ELEVATION [)3 -03 E.3-E:3 TYP INTERSECTION or TYP NOIIIZ fiPOOYE TYP YERT 61/00YE
£L£VAT/Ol*I J.J .. J,J WATTS BAR YCRTICAL ANO SCALE 3/4"rl**Cr FINAL SAFETY ANALYSIS REPORT D£T.4/L A3 D£T.4!L 83 NOT T05C4l£ 2 REI}() MIi i/NiT sCAu:: 3/4"*r-o* AUXILIARY BUILDING UNIT 1 CONCRETE NORTH STEAM VALVE ROOM OUTLINE TVA DWG NO. 41N397-3 RC FIGURE 3.8.4-48
t!) z
- s a
- :
0 0 w z ~ < EJ. 781.0 I i ! i I N c_£'L 781,0 n.Te,.o. 1- I z I I !
! iI
~ _L£Lm12 I I
~EL777./,l
,:OPT,ONAL. CONST JT c_OPTIONA/. CbNST 0
< £i.775.5 EL 775.~ n. 775.S;;, I t.)
@- DEfAIL 83 rEJ.770.25
~~
EJ. 770.25::::,. .J I I I I I I I EL=-1~1~0-=zs"-"--+--l--------- r----7 cCONST JT I I I ' I e,CONST JT CONSTJT 1 EL 768.0 n, 768.0 s 1&a.P I I I I iI ,,__OPT VERT CONsT JT IE£ 75.3. 0 TO EL 781. o ~ I I 7'-0" I
! ! rEJ. 7,3,0 I n.7&!.0--..
2-41N712 - 2-41N712 MU{ flNT 2 lWLYJ; MINIMUM ACUTAL STR£mlTH OF CON::RETE IS
.f600 PSI IN CI.OIJOED ARf:.45.
NQTf if'itHV QII Yl* Mlll!HI ACl/00. Sll?ENGTH OF CCNCR£/F IS 4600PSIINCIJJIJD£I)ARE'AS NOT£: FOR NOTES AND ll£F£ll£NC£ ores Sf£ 2-41NJ97-1,2. BEEEBENCE PBAWIN@
-41N10058-4B---CONCRETE POUR DRAWING 41N1005B-4C---CONCRETE POUR DRAWING pcGROOYE II * ~-1 I .r I 1-'* 4" u<->K-i lJ
"" ,1" f-:1 I I.J_ l:..tJ1:_
C()l,/$7 ..n:---G I/ °'! I 111 I ~ ~ AS REQO ~--:,t r--\.1!1./:.n F< I >9 I .'\~ I *~**--1*1 I I I I I l I' l I* I I* b
- I 1
1 (/YW.sr.d .4,:;~rm FOR LOCATION SL"E BJ* BJ 1.2 NW {UNfT Z QY/y)* TEMP CONST OPNS.../
$"/ .VATll'>AJ IJNMIAI ACt/TAL STRENG7H OF 03-0.s
?VD 1..1,f'jD,., ~"1il'Vlvr ~~~!!v.vr CONCRETE IS 4600 PSI IN cwtJJJfI) _ , ,
ELE/IATION J.J.J,J OF YC/lTICAL AND IIORIZONTAL GAOOYC$ S&ALEJ**r-o- WATTS BAR FINAL SAFETY D£TAiL A3 DETA;L 83 ANALYSIS REPORT NOT TO $CAL£ 2 KEQD PtK t.JN/7
$CAL£ , ....,.,_O ..
AUXILIARY BUILDING UNIT 2 NOTE, FOR HATCHED AREAS SEE UN IT 1 DIG 41 NJ97-3. CONCRETE NORTH STEAM VALVE ROOM COMPANION DWGS, OUTLINE 2-41W397-1, 2, &. 4 TVA DWG NO. 2-41N397-3 R2 FIGURE 3.8.4-48(U2)
ITEM QTY DESCRIPTION W.TL NOTE
- t3* IS GOOD FOR SPECIFIED DUENSIO.S 1 Pl 1/rxs*xLENGTH BY FIELD A36
(.!) EE z TE
- a
< DO NOT WELD 0::: 0 0 4*-0** r,:, w z ...... ,v
~~
O'V YEflTIQIL l..E£i AS I- INOTE' IIEQUIR£0 TO FACIL/lil/TE I* z LOCATION OF ALL CCWSTIIUCTION
- Ii II MAJN"r,IJNN6 I"'
ANCHORS n, N t1~, EDGE 0/ST;IJ,I/Z MIN 0 I MK/ < 11! ~ fS*~*ttrrPJ I RcQ'tJ tJ** l'O" u l.. I 1J**NY NOTE: ITEM 2 MAY BE TRIMMED AND MK 3& MK6
~ SELr-CLOS/N6 SATETY ~ WELD MAY BE ClilITTEO AT Alli$- I! 11EQ'o GATE 1rff.U#&"956*S)~
()/ i ZI (6 PLACES) EDGE OF W8 FLANGE TO OBTAIN NECESSARY CLEARANCE Alli fi- I REQ'O tf'*l'*O'
.JI-JI TYl'(IJ.ON.)
- '.J "i J"*/'-O~ EXJSTJNGL4x4 TYPTOPAND ~ -
BOTTOM OF ANGLE 2..._ ** CUT GRATING EDGE f6 L5x3 AS REQD
- . - Tr, rlt:Aor,ot:ll/ll/:,
TYP - ~ ~ - ~'MAX TYP.
----__...,....._,,...'"ill---'----23/4 0 (MIN}
- HK STIFF PLATE TYPICAL 6 PLACES CUT FLUSH WITH EDGE OF BOTTClil FLANGE DET W1 iVvO QA txa/M&,T,r7iON REQ{ltl/£0 ~ HANDAAIL.
,'.3. fi"a!? OJ1EllS/ZEl) 1-/C!LD ,'If SI-la; ~*f~S, t:!EUJ 1,1,,-yUSE C"txl nAIE WASHUfS WITH A Fil.LET 4* wa.o tw ~ .5!!)£S (.-HI-¥. z ~SlTE SlaS).
- 14. ,£Los SHOWN ARE MINIWAI SIZE ANO LENGTH KEQIJIHED. ADDITIONAL WELDS NOT SHOlfN ON ORAi/NG M4Y BE USED AT THE DISCRETION OF THE FIELD ENGINEER. NO OOCUAENTATION OR INSPECTION IS KEOUIKEO FCW ADDITIONAL lfELOS.
- 15. THESE PLATFORMS ,ERE EVALUATED OUK/NC THE SEJSUIC/CIVIL EVALUATION PROGRAM (/fCC-1-1'#19). ACCEPTANCE IS PROVIDED IN CALCULATION /fCC-1-1316.
WATTS BAR FINAL SAFETY ANALYSIS REPORT AUXILIARY BUILDING UNIT 1 MISCELLANEOUS STEEL NOTES (CONT)
- 18. TVA FABHICATED SELF-CLOSING SAFETY DEVICES AKE DETAILED ACCESS PLATFORMS ON DHAlfINC 48£956-5.
- 19. PREFABRICATED SELF-CLOS/NC DEVICES APPHOVED BY SIT£ ENCINEEHINC INCLUDE FABENCO'S SELF-CLOS/NC SAFETY CATE MAIN STEAM VALVE ROOMS OR THE INTHEPIO INDUSTHIES BAR TYPE CATE. OTHER SELF-CLOS/NC SAFETY DEVICES APPROVED BY SIT£ £N(;/N££RINC AND TH£ SAFTEY NOTES (CONT)
- 16. ALL N£lf llfSTALLATIONS SHALL UTILIZE SELF-CLOS/NC SAFETY DEVICES (GATES).
DWG NO. 48W1211-1 RJ ENCINEEH M4 Y BE USED.
- 20. US£ CARBON OR STAINLESS STEEL SELF-CLOS/NC SAFETY DEVICES INSIDE TH£ REACTOH BUILD/Ht;.
- 17. EXIST/NC SAFETY CHAIN IN NEED OF REPAIR OR AS SPECIF JED BY TH£ SAFETY ENG/NEEH SHALL BE FIGURE 3.8.4-49
t, z
- K, Ct
C, C, w z 1-z C, < N7!FdNIT 2 fS(H()SllF HINJ FrR All:M° lETAIL tJ N71FdNIT 2 IS fHOSI7F HWJ RR AllM° fETAJl. IK1J ID'L 5XJX)" ALTIETAIL Bl N1S (222227' ~** l.f. llEZ1)5 S10IIV A/iE" MINIMN SIZ£ MO LEIC/1/ IIEWIHEll. AlDITICIVAI. IIEZ1)5 MJT S10IIV av ORA/flt& UIY EiC IHl) AT lJC 0/SOiUia-l tT lJC FIElli a&.IAEER. MJ OX'IA£NT,1,TJ"a-l a? /NSPECTiav IS /IEWiflEl} RR AIDIT/m41. IIEllJS. 15.H4/\t:WA/LS. SELF aai/AG G4TES. 61i'AT/AC, KIO( If.ATES. lAaJERS. MO ASSX/AlEJ H4RlJ/fAH£ UIY EiC fflXUiE/J MO FAHf/CAlEJ a:1-FCliM/AG llJ 5£/SII/C 11/T aASSJF/CATiav (TVA CA/RZll'Y l(L). JLD[F[D4T/Q( IQTfS,* I.ALL 57FEZ(A36) /JllESSOTfEli'll'/5£ MJlEJ. 2.S!RlC!l.ffAI. SlEEI. £RXTJ"a-l 5'i4/.l EiC l'EliFrRIElJ IN A£22l'ill4Jla" If/TH liEl7-/7El l'litXElXR£ 2 ~ 1 *
.J. ALL IIEW/tc 5'i41.l EiC 1'£/rC1i1Lll IN A£22l'ill4Jla" l'/lH liCCHlEI.
5"'f 25KJ2--<XXJ-(M(-6JI-0X01.
-I.PAINT/AG MO (XJI/T/tc 5'i4/.L EiC I N = lfllH liCOfTEl l'fitXEXfiC~ -
UFSAR AMENDMENT WATTS BAR (222227' FINAL SAFETY 1.#vIT 2 SEE OJI(;. 2~/Yl2 fl-7 ANALYSIS REPORT AUXILIARY BUILDING UNIT 2 MISCELLANEOUS STEEL ACCESS PLATFORMS MAIN STEAM VALVE ROOMS FOR HATCHED AREAS SEE DWG NO. 2-48W1211-1 R3 UNIT 1 AC DRAWING 48W1211-1 FIGURE 3.8.4-49(U2)
Figure 3.8.4-49A Deleted
FIGURE 3.8.4-49B DELETED
t, FOR DETAILS SEE 48N710-3 CTYP) z EL 803 '-5" EL 803'-6 1/4* 3 TYP 5 VENTS < \ BLOWCFF ROOF TYP :i VENTS a::: c:, ~ EL 803'-4 1/2* c:, w z 1-z
- i;;
c:, u I I bl 111 1.,, I I I NOT TO SCALE I I I NOT TO SCALE BLOWOFF ROOF SEE 48N711-3 FOR DETAILS I I 1S, "SOME WELDS SHOWN ON THIS DRAWING (SERIES) WERE FOUNJ Tn 1::1.1: ni::i:1r11:tJT IIY Tl,111:' lUl:'I n ~ *
- I LI.T'V'l,l,I Dl::ll"l.11:'rT, l:'V.1.1 I llt.Tl:'n BY WATTS BAR ENGINEERINCi PROJECT AND FOUND TO BE SUITABLE FOA iSEAYICE, SEE DESIGN CALCULATIONS FOA EVALUATION",
111:P'i:'P'lil: TO RIM41. NO- ,-,:,~1tanQi"JH1;1 Q .UJn -'-h~ ~R WRN MP'l:UtMQ_ 17 .ALL MAIN STEAM SAFETY VALVE MARK NUMBERS ARE RUNNING LINE TRAP LA/ II INTERCHANGEABLE AND MAY BE INSTALLED IN THE SYSTEM AS REQUIRED FOLLOWING REFURBISI-MENT. SEE 47W-4-15-2 FOR PIPING DETAILS Pi.AN £i. 758.0 t UNI Tjj AS SHOW/{ REFERENCE DRAWINGS: UFSAR AMENDMENT 47SM415-ISII.I. OF MATERIAL 47W8O1*/ Fl.OW DIAGRAM-
;,;;,.;;; t REHEAT STEAM WATTS BAR FINAL SAFETY RELIEF VALVE I 'NIT J ANALYSIS REPORT l.OCPI 1~-03;7
,e:-031e IE.-0354 POWERHOUSE S£CTiON Ai*Ai NOT TO SCALE LOOP, 1£-0.!SS NOT TO SCALE UNIT 1 LOO, 3 1! - 0J96 IE-0397 EXCEPT AS NOTED MECHANICAL MAIN STEAM RELIEF &
SAFETY VALVE VENTS TVA DWG NO. 47W415-1 RK FIGURE 3.8.4-49C
t!) z ,_, FOR DETAILS SEE 2-+8W71D-7 Al+ 3 20*-o* EL 803' EL 803'-6 1/+' a::: BLOWOFF ROOF TYP 5 VENTS TYP 5 VENTS c::, c::, 5°-0* w z o**". ***1** T/S 800'-6* 1-
...... ... I z
,_, ~
- i;; ! 6 c::,
(.) 11 I MARK NUMBERS 1140 MA.ll I I .;*.,.,.,IJ I#
,.,..,,,.. ,..,,~,,
1"4Hd 637 9214 ARE TYPICAL FOR 20 SAFETY VALVES I TEMS A THRU T MARK NO.S 01/f. MU/MY a* ,,,,ntwrr~J
~ Uflrl ,_,t pi/U. ARE PRECEDED BY 471400.
I I I I I I I _/'1~4*i,1,~pipw NOT TO SCALE I I I I I I I EL 782'-8 1/2" NOT TO SCALE EL 782'-7 1/2" EL 782'-r BLOWOFT ROOF SEE 2-481711-6 FOR DETAILS
. I I I 11 SEE DETAIL 81 I FOR DRAINPOT I I 2-PCV-001-0012-T PCV-001-00D5-T11\-
2-PCV-001-0023-T 2-PCV-001-00J0-T I I _0;--~-,J~ -0
- 16. ALL MAIN STEAM SAFETY VALVE MARK NUMBERS ARE INTERCHANGEABLE AND MAY BE INSTALLED IN THE SYSTEM AS REQUIRED FOLLOWING REFURBISHMENT.
SAFETY CONSOLIDATED 3NC203+ P09096
-~~
j.*.*:. r-.-.""*,,"'*T"O""-------+--------------,-:-,.-=-..-:--:.: .. :.:i Cl 7$7,0;
'1 * .. * .
"I~
d
~
~
I CLOSED DRAIN RUNNHC LINE TRAP SEE +7w+15-2 FOR PIPING DETAILS I L...,41 PLAN EL 758.0 t UNIT 2 OPP HAND a. NOTED SEE SAFETY VALVE TABLE F1 FOR ITEMS A THRU T NOT TO SCALE EXCEPT AS NOTED DETAIL C1 TABLE F1 {SEE NOTE 16) SOUTH VALVE ROOM SAFETY VAL VE SET PRESSURE HEM SAFETY VALVE UNID (PSIG) A 2-SFV-001-0522 1224 B 2-SFV-001-0523 1215
,,' C 2-SFV-001-0524 1205 AMENDMENT
~ .. D E
F G 2-SFV-001-0525 2-SFV-001-0526 2-SFV-001-0517 2-SFV-001-0518 1195 1185 1224 1215 WATTS BAR H I 2-SFV-001-0519 2-SFV-001-0520 1205 1195 FINAL SAFETY L RELIEF VALVE SU/:PORTS I J K 2-SFV-001-0521 2-SFV-001-0512 1185 1224 ANALYSIS REPORT UNIT2 L 2-SFV-001-0513 1215 LOCl'I l!*lll/7 M 2-SFV-001-0514 1205 Ze:*03!e N 2-SFV-001-0515 1195 REACTOR BUILDING SECTION A!*AI I.COP .2 ZE-oJ54 ze-a,5!5 0 p 2-SFV-001-0516 2-SFV-001-0527 1185 1224 UNIT 2 NOT TO SCALE lOC9 a 2E-C396 Q R 2-SFV-001-0528 2-SFV-001-0529 121!1 1205 MECHANICAL 2E-OSl7 s 2-SFV-001-0530 2-SFV-001-0531 1195 MAIN STEAM RELIEF & SEE SHEET 2 FOR PIPINC T 1185 SAFETY VALVE VENTS AND HANCER LOCATIONS - TVA DWG NO. 2-47W415-1 R1 FIGURE 3.8.4-49C(U2)
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-50
SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 FIGURE 3.8.4-51
Cl z 3 18'-0" <( "I r OC-TYP D::'. 0 TVA SPRAV WATER CONNECTION 0 w z <( f- --- t Screens t) -COOLING z TA. TOWERS <(
- a' Top of wall ( t Screen 0 El ?51'*0'
<( u 6 pieces at 5 1-0 11~ 2 {Jleces at 2 1-6 11 PUMPING STATION DECK PLAN EL 741~0" insfollsd between El 123'-6" und EL 74!'-o" for screen guMes for scree11 IA- A, NOT TO SCALE Weldfld in accordance with welding Ml( I procedure No. SM-U-1 or GM-FC-U-1.
~I IP. f, typical; rSym I
about INTAKE PIIMP/N/J _,, .-, ASTM A588 TYPICAL PLAN
~ -
MK -34NZ/O-/ 4 RE:0'0 r!." rf-1 Strip anchor 1j" x f" CI/ICl<AMAUGARESERI/~/ / 3" x I II L - bar, located HEY PLAN on 18 centers. ASTM A-570, NOT TO SCALE grode D. I
~ ~I=::: 3 ..~~.I--.L---l-----l.J::===-1==-
For Manu-fa.c-turer:S de.fa.ifs .see Envire.x .Ina. File, TVA con-t-ro.ct- 74CSa - 85143.
-2.
/6 t Screens 2 SPEED DETECTOR SWITCII DETAIL A PROVIDE OPENING WITH REMOVABLE INDICATOR BY TVA FIELD WATERTIGHT COVER IN HOUSING AND NOT TO SCALE HOLES TIIROUGI/ SPROCKETS TO t light fixture PERMIT ACCESS TO END OF SPRAY mount /lor,zontal!YJ), PIPE FOR CLEANING. NOTES:
Four scr~ens required. Capacity of' eac// screen-25,CXJO 9pm of' clear wafer WATERTIGHT HINGED DOOR, BOTH /'or clean screen with minimum headwater El 666*-o* ENDS, LARGE AS PRACTICAL. and mu water velocity of 20 fps. A,aila/Jle spra_y waf,,r per screen to /Je li'O 9pm ,sf
/lii,ge-d1 ren,oyable TVP 80711 SIDES 3"-150# WELDING 90 psi at the nozz /es.
low~r section NECK PIPE FLANGE, RAISED FACE, Control sfqtion furnished by TVA. WITH ONE BLIND FLANGE, GASKET of rear housing BOLTS FOR EACH SCREEN, BY CONTR.
~ EL 742'-2"-.___~---- QUAU7Y ASSURANCE 5'-2"+/- 3/4" 1. SCREENS ARE CATEGORY 1 EQUIPMENT.
GRATING BV TV A i'Yidth of well 2. CONTRACTOR REQUIREMENTS IN ACCORDANCE WITH Mor far to wit TVA SPECIFICATION 2023
- 3. FIELD REQUIREMENTS, 4" f'langes on sluice &P)*
Corrosion r,s/sf4nf *rram~ n-----+-----_.,,, = A. FIELD TO INSTALL SCREENS IN ACCORDANCE WITH ACCEPTED SCREEN ERECTION PRACTICES. Max water at screens anchQr bolf..s. by scr~en B. FIELD TOUCHUP OF PAINT IN ACCORDANCE WITH El 737'-6" confr.
\Guides fbr descending clia,nsry CHIEF ARCHITECT'S INSTRUCTIONS.
- 4. ALL MATE RAILS BY 71/A FJELD AND FIELD WORK SHALL BE INACCORDANCE
_£Max 11ormal water El 683'.0" WITH GENERAL CONSTRUCT/ON SPECIFICATION N3G-B81,
~ QUALITY LEVEL .II.
I Cc,sf iron guide b_y ! M~~ h*ight of frame screen contrdcfo_,_f------l!f------ T bu ,sj <i_ screens SCALE: NOT TO SCALE EXCEPT AS NOTED
.s~,hons/5'-0" ~ __FM1il normiJI waler El 675'0" Except as noted '-------j-"t!.~f-~~~~~-d-=g:c~~~~=fi-Jl--fl ,._'i "'"de
'-- V in DETAIL A. cont, Flow
[M,~ wal*r wilh loss of
~risfream dam E/666'*0" r
C 7C ~1 h: TVA FIELD TO LEVEL FOOT FRAME BY WELDING SHIMS UNDER CORNERS AS REQ'D.
~ L____4_'-_o_*_ _ _ _,.
SHIM AREA (LENGTH & WIDTH) FOR FOOT W,d/11 of tra_y FRAME TO SUPPORT ENTIRE SCREEN WIEGI/T, Tra.JS,, c/2a111s 1 and sprocJrefs SHIM THICKNESS FOR SMALLEST GAP BETWEEN FRAME & CONCRETE. ADD STRIPS TO FRAME IF NECESSARY TO ao t s/Jowr,
~1 UFSAR AMENDMENT 3 CLOSE GAP TO LESS THAN f" -
Fl.El. 652'-0' 1 c-c NOT TO SCALE WATTS BAR FINAL SAFETY SECTION A-A B-B ANALYSIS REPORT WATER SUPPLY UNITS 1 &. 2 INTAKE PUMPING STATION TRAVELING SCREENS ARRANGEMENT TVA owe NO. 34N210 RC FIGURE 3.8.4-52
c.., z c:, c:, w z 1-z
- i;;
El)(;E OF P/11,/P/NG STA c:, (.) 6A
,/H ~
~
,li A I METHOO OF' FXC.IVATIO#'
UCAYAT! //OCI< I# A SAW- TOOTH MANI/I// TO A /C/NJL LOCATION, A MINIMUM OF' A INCHlS 8EYOND TH£ 01./TUNI SHOWN F'O// TH£ STIWCT/1//AL CONC//!Tl, Pt.ACE A Ct!HC//IT£
.$1/8/JOU// 7'0 lUYAT"IONS Ai'/0 O!MlN$!0NS $HONN.
CONC//rrl Ft/JI SU8POUI/$ $/IA" 8l CLASS 201.I anr i OJI 201.I AFW IN MCO//flANCl WITH CONST//UCr!ON snc1;IcATION G*2. A .1l F'O// THI Al/EA #TWEEN TU //£7"AINIH6 WAlU UCA//llrE //OCK TO FINA/. i.OCATION A MINIMUM 0,t' 6I/ICHU 1£LOW TNl E'UYATIONS SHOWN _, ,.,._ ___T/1£
,._.,_. ON /NTAI(( CHANNEL
.., ,.,.,.,.,...,_...,_ fl//AWINQ
,u_,< ,._,. ..,,.,,
NTS
/0. FllTER MATCRIAL SIIALL BE IN ACCO//OANU WIT/I
$P£CIF'ICATION Tl, $£CT/ON 107$.
ii. WHt://E GROUP$ OF' //ARV/NG BAIIS A/IE E!ILL£D FOIi TWO FACES, ONE-HALf" or EACH .i.ENGTHS 8ii.LiNG iS PLAC£D IN EACH F~ce. A It. TO FAC/LITATl CfJrSTR//CTIQN CLASS ;l{)I.SA FW Ct!NCRE7E ;VAY M S//BSTITVTEP F/JR CLASS 301.SAFWCMC//lrE. WATTS BAR FINAL SAFETY ANALYSIS REPORT WATER SUPPLY UNITS 1 &. 2 C3-C3 t DJ-D3 CONCRETE 03-03 OPP !IANO NOT TO SCALE RETAINING WALL AT INTAKE PUMPING STA OUTLINE&. REINF TVA DWG NO. 31N224-3 RE FIGURE 3.8.4-53
~U/IIBEDMD PUT(
ll(JtUH MWAUTE. CPNC.
\__ Cl El ?OJ.75 I , FIJ6 IJUAll..t SE£ MlMtJ
~ R.ld. PIEK(£ T/17: I. NMTNEI/N JR.
DI-DI
/NT,1/1(£ PUMPING S TATIO? L~::::;::;;;~~~~::~~~~:::;:~~::::;:::::::;::~===~;~~=:::::::::::::: ~ MTEP FF4 n, lnT El-El NTS NTS
_,___,,._+.--t------....!:.2!1"-='-..s0:..."___~--:----1::-f-~=:::lf--;-----L---L.J!i Bl 1cue DV(T I ,r--~--~-----;----,-+--rJ.-1 1(10'/lf;.N1. I i ---+---~----------- II ------- ] KEY PLAN L H-"-i'--'--P...._:...._.:.....+~,.__....._,..,_..y.,._f-..---1...i,;,,..-f-*-,-X-P_J71.,1M-~-,l...---+"'==--;,,,-i i FM SPACING c!,.. r,r RotJJ,' ! 1
""1 Bl-Bl NTS
~ 1 T.:':1-~--'<=-=l=7'=W=3_0_0_*_9~~C~C:~L~7I_O_.o_ _ _ +-,
l
'i -----+*=====+-, 1*---=""'""'="' I I N()TES:
NTS NOTE fl: FILL 114*30 AT EL 710.0' l"ER SECTION Cl-Cl: ALL a:teltUE Mi\TEIUALS AND IORKNANSHIP SHALL IE IN ACCORDANCE IITH THE LATEST EDITIONS <1f ACI 311 All> ACI 301. ALL L BE CLASS 300.75 Afl MIN, (3000 l"SI Olt BETTER) AND IN ACCOIIDANC[ 11TH C£NERAL EICINEEltINlil SPECIFICATION C-2.
,. LL BE GRADE 60 DEFCRIED BAltS CONl'Dll,IING TO ASTM EEL SHALL BE ACCURATELY PLACED 11TH CLEAR Cl1/Elt IN AOCCIIDANCE 127 '-o* NORTH EAST WALL
- ANO ADEOUATtLY SUl"l"Oll!Tm IIEfOltE THE coteft[T[ IS l"LACED.
115'-0" SOUTH WEST WALL SECTION lA-lA FOR ADDITICffAL INFORW.TION. (CC:W-.CRETE FILL - SEE NOTE f1) NTS T~ Of CONC EL 710.73 NTS NOTE G1:
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REGU!!UD. UM!T Hf/6NT Of" !..!Fr ro 14 FUT OR USS. THE LOCATION ANO ANALYSIS REPORT NVM6U OF JOINTS TO BE OETE/11'/INEO a, nt,o.
,Cl1Al.t;T .1r AUXILIARY BUILDING I EL 1u:s UNITS 1 &. 2
";.: . I 't,-- ,</K370*3 -fL 'Mil/ES CONCRETE
~1._1_1 I I 4A CONr ! l2"4C. £F SCALE ;f**t'*O*
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UNITS 1 & 2
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~ - . . . . . - - - - ~ Cl 7Sl0 PARTITION WALLS OUTLINE & REINFORCEMENT TVA DWG NO. 41N373-2 RJ FIGURE 3.8.4-61
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.1 PARTITION WALLS OUTLINE&. REINFORCEMENT TVA DWG NO. 41N373-3 RJ FIGURE 3.8.4-62
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IN SECT D4-D4
.SHIM A'. /x,3fx FOR EXCEPTIONS 10, All /JOLTS ro 8! ASTM A307 ASTM AJ~ T!IREAD£D
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;( ;( ;( NOTE: OPENING LOCATION 17\ FIELD MAY CUT CONCRETE TOLERANCE e* +/-. (J) 1-4A VERT BAR, NF & 1-4A DIAGONAL BAR, NF WERE OPENINGS IN THIS AREA \:../ 2-4A VERT BARS, 1 EF. PARTITION WALLS OUTLINE&. REINFORCEMENT TVA DWG NO. 41N373-5 RL FIGURE 3.8.4-64
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(.) NOT!' FOR ADDITION NOTES AND REFERENCE DWGS, SEE 10N203-1. SETTLEMENT STATIONS 34 THROUGH 41 SETTLEMENT STATIONS SS-34 THROUGH SS-41 SHALL BE SET BEFORE
~
INSTALlATION OF TURBINE GENERlfTOR PARTS. EXTREME CARE SHALL BE TAX.EN NOT TO DISTURB THEM DURING TURBINE AJ,JD GENERATOR 5$*11,lL 7.11.0 SS*J,lL 7.11.0 ERECTION AND STARTUP. ELEVATIONS SHALL BE READ C3 (TOP OF WALL, TY...;P)--~--',-.--;,c,,--O'T'---*fff'tt-- =-...,,.M------+-....:)1(.,.-----,~--+alr-'-(70!' or WALL, rr,) ( 1) IMMEDIATELY BEFORE INSTALJATION OF TURBINE-GENERATOR PARTS (2) IMMEDIATELY BEFORE INITIAL OPERlfTION OF THE UNIT, AND r .. ' l SS*1A,EL 743.744 \
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NOTE: HOLES CUT FOR NEW ANCHOR BOLT ASSEMBLIES lrfl( 1 MAY BE TORCH Cl.IT FOR PANELS MK. 3 ;_ 9 1 & 10L. HOL!S FOR ABANDONED BOLT ASSEMBLIES NEED NOT BE REPAIRED. 6TC*PS,TYP) ... R\ B=IJ .tr I MK,. aNLY
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"* CATEGOl!Y !FOR THE ECN 347!1 REVISION OF DRAWINGS 44N330, -
44N331, AND 44N332 L , !I I 10. TVA FIELD TO SEAL WELD ALL THREADED ~NECT.ONS. II. FIELD HEQ(/11/EMENTS; WEl..lJIN& Tl/ lie IN A<<MOAM:£
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! ! I !II I l ,~ _.--Ml( 44N332-19 1
ANALYSIS REPORT NOTE1 ALL DIMENSION TOLERANCES ARE /
/ I I' : !i 10) I 2 I 112"P<ATE IWtftl/."c,,coNDA t 1 /32" L.i ~ I/2'1 I 2 1,1**~SOLT I ANSl,818.2.1 DIMENSIONS I I 7£ UNLESS OTHERWISE NOTED.
I I I' I'-=-' I 10<<~,c 1M****o*M*-""*~u AUXILIARY BUILDING I! I I(s) I 4 1~1!1~z,~,.!l.irl/4"I fi'ir~4f) UNITS 1 &. 2 I!/ 117' ;;,:.w HEX NUTS IIDIM. 11.:.1 II ** II3i4"IO ASTM Ao** TYPE 030 PER ANSI ., *. 2.2 SPENT FUEL POOL FUEL POOL GATES ARRANGEMENTS TVA DWG NO. 44N33O RG CN.T .S.) FIGURE 3.8.4-69
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- 3. AOO TWO INT£RMEO/ArE GU/Oc VANES TO EACH S/0£ OF' THE IJArES SHOWN REMOVABLE 12
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< TRIM 2* ~ ~~ 2 * (..) LOWER GU/0£ VANES) AS SHOWN IN H-H ON 44N 331. Sl(JNt 8-8
- 5. AOO INrERM£0tAr£ GATE GUIOES TO rH£ WALLS EIGHT Pt.ACES CUT our AT SOUTH WALL w
TO MAr£ WtrH TH£ INT£RMEOtArE GU/0£ VANES WHEN THE L6xtixj* GAr£S AR£ IN TH£ OP£RArtNG POStr/ON.
- 6. ALL MATERIALS ANO W£LOING TO BE IN ACCOROANC£ WtrH 304, L SS FERRULE QUALtrr LEVEL .rr OF' CONSTRUCrtON SPEC/F'ICArtON BY SEAL SUPPLIER Gl1T£ SVPPORT N3G-88/, ~* AT WEST WALL, TRIM TO BE EVEN 2 REQ'O. STA/NUSS STEFL - Ml< 6 j / I WITH TRll,MED GUSSET (8-B)
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P.JOTES I. FOR GENERAL NOTES SEE 44N330.
- 2. ALL MARK NUMBERS OETAILEO ON THIS DRAWING TO HAVE PREFIX 44N332 UNLESS OTHERWISE NOTED.
- 3. SEAL TO SE TESTED AFTER ASSEMBLY TO A PRESSURE
..-s.s.,tf OF 40 PS!.
- 4. ALL SLINGS TO BE DES!GNED TO ANS! STANDARD B30.9*I971.
911ALITY ASSURANCE NOTES:
- i. MK 7 TO BE PROCURED UNDER A QUALITY ASSURANCE PROGRAM L 4)(3,cJ*or AS DESCRIBED IN TVA REQUlS!T!ON, equlval~(u.)
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- TRAl../S~E~ CANAL ANALYSIS REPORT I l PCOO W/0 A!)()!() PC*OPl'O$!TE U...dNO*MK 5*T.R.A,N~P. CANAi:
5E.4L NOTE: AUXILIARY BUILDING 3 12.Eal)
- MK I THE TRH,.. ING OF THE GUSSET PLATES IN SECTION C-C ABOVE INCLUDES ALL THE GUSSET PLATES ASSOCIATED WITH THE ( 2)
LOWER GATE SUPPORT GUIDES ON THE CASK LOADING PIT GATE ON UNIT 1 SCALE 3/4" = 1' WEST WALL Of THE SPENT FUEL PCXll. REF MK 19 ct. MK 20. SPENT FUEL POOL COMPANION DRAWINGS, 44N330, 331 FUEL POOL GATES SEAL DETAILS TVA DWG NO. 44N332 RF FIGURE 3.8.4-71
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1- i-!EAP CAP z 2. REMOVE THE EXISTING MIDDLE GATE RESTRAINT GUIDE 1'4NES FliOM GATES. 01.JTIIOI!! $/!AL lleTAIN~/¥ ~ 3. AOD TWO INTERMEDIATE 611/DE VANES TO EACH SIDE OF THE GATES AND FIVE ADOIT/ONAL CROSS MDIBERS TO EACH SATE AS SHOWN 51-'0WN /ll!!N/OVAaLe Cl IN THE POOL SIDE ELEVATION DH DRAW/NS .UN33/. i,, (..) "* ~f#of/!fFf:l'/, !t,'lfE T~/'lEr'!f$f!Dtifl1Rf tft,[#EA/ITES LOJYER SU/DE VANES) AS SHOWN IN H-H ON 44N 331.
$. ADD INTERMEDIATE GATE 61/IDES TO THE WALLS EIGHT Pt.ACES TO MATE WITH THE INTERMEDIATE SIi/DE VANES WHEN THE SATES ARE IN THE OPERATING POSITION.
- 6. ALL MATERIALS AND WELD/NS TO BE IN ACCORDANCE WITH 3041. 55 FEtt/JLE fJIJALITY LEVEL ]I' OF CONSTRUCTION SPECIFICATION 5Y SPAL SI/PPUU GIIT£ SVPPOR T N3G-88/. Z REQ'D - S1AINl.£SS STEH -MK 6 m,)j t,:')
~ GENERAL NOTES SEE 44N330.
- 2. ALL MARK NUMBERS OETALEO ON THIS ORAWING TO HAVE PREFIX 44N332 UNLESS OTHERWISE NOTED.
- 3. ::,.CAL IV DC. 1c.:,1c.u Al" 1c.n: ~~c.,,n:H.. T 1'1,1 ""~~\,Inc.
OF 40 PS!.
- 4. ALL~ 1 1 ,..~ m 1¥' ni:-c:.1,u.1i:-n m 4.t1.151 ~T.ti.Nna.An RVl,Q-IQ71, f.'Mgll; W"J!M't"N~=j:unuc.n ~ ~ ~unAnvc. ~ruw AC! l'\C'C!,...OICIC'I"\ l"-1 T\/11. 0i:-n111c:.1T1n~.
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SEAL AUXILIARY BUILDING
- ! !!ECJD * ,,.,,KI UNIT 2 St:.I!/~ ,t"'.-/!.O"' SPENT FUEL POOL FUEL POOL GATES SEAL DETAILS TVA DWG NO. 2-44N332 RO FIGURE 3.8.4-71(U2)
Figure 3.8.4-72 THRU Figure 3.8.4-80 Deleted
WBN 3.8.5 Foundations and Concrete Supports 3.8.5.1 Description of Foundations and Supports 3.8.5.1.1 Primary Containment The primary containment foundation consists of a 9-foot-thick circular reinforced concrete structural slab, measuring 131 feet 7 inches in diameter. The outer 5 feet, where adjacent to other structures (225 of the slab), is thickened to 16 feet, while the remaining 135 portion is thickened to 12 feet for the outer 13 feet. These deepened portions are transitioned upward on a 2 to 1 slope to the bottom of the 9-foot-thick portion. The slab is keyed into rock in the central portion by the 8-foot-thick walls of the reactor cavity extending a total of 26 feet into rock. A 3-foot-thick concrete subpour underlies the structural concrete and caps the top of the irregular rock surface. This serves to preserve the rock in its native state of being under pressure, thus preventing deterioration of the rock surface. The base rock consists of interbedded shales and limestones. See Section 2.5.1 (historical information) for additional discussion of the rock base and foundation treatment. The interior concrete structures described in Section 3.8.3 constitute the support system for all equipment in the containment structures. All major equipment supported on the foundation (steam generators and reactor coolant pumps) is anchored through the steel liner plate into the 9-foot-thick concrete base slab, thus preventing the liner from becoming a stress carrying member. The base liner plate is anchored to the foundation through the use of embedded 'Tee' shaped steel sections which have provisions for leveling before concrete is placed. The embedded anchors are used as screeds during the placement of the concrete to ensure that a flat surface is obtained coincident with the top of the anchors. All welded joints in the base liner plate are made at anchors. All joints in the base liner plate are equipped with leak chases to facilitate testing for leak tightness. As defined as part of the primary containment, the concrete structural slab and its liner are subject to the inservice inspection requirements of ASME Section XI, Subsection IWE, Requirements for Class MC and Metallic Liners of Class CC Components, and Subsection IWL, Requirements for Class CC Concrete Components. However, the concrete structural slab is inaccessible for examination as it is covered by the liner as described in Section 3.8.2.1.1 and the containment floor structural fill slab as described in Sections 3.8.3.1.1 and 3.8.3.1.2 and is exempt from examination in accordance with IWL-1220. However, the concrete structural slab is subject to the repair requirements of Article IWL of ASME Section XI. The inservice inspection requirements for Class MC and metallic liners of Class CC components are contained in Section 3.8.2.7.9. 3.8.5-1
WBN 3.8.5.1.2 Foundations of Other Category I Structures Auxiliary-Control Building and Associate Structure All of the Auxiliary-Control Building, except the waste packaging structure, and the condensate demineralizer waste evaporator structure portion is supported by a reinforced concrete slab placed on a 4-inch-minimum-thick concrete subpour which caps the top of the irregular rock surface. The Auxiliary Building portion of the base slab is 7 feet thick while the control bay portion is 5 feet thick. The entire base slab is located on three different levels with continuity between these levels being provided through thick walls. The thicknesses of the slab were selected primarily to provide sufficient rigidity to minimize differential vertical movements of columns and walls and secondarily to reduce shearing stresses in the slab itself. Due to the thickness of the slab, anchorage into rock was not required to resist hydrostatic up-lift pressures from maximum flood conditions. The waste packaging structure is separated from the rest of the Auxiliary Building by 2 inches of fiberglass expansion joint material. The 45-inch-thick base slab at grade Elevation 728 is supported below Elevation 725.25 by crushed stone backfill placed in 4-inch layers and compacted to a minimum of 70% relative density. The base slab of the condensate demineralizer waste evaporator structure is 2-feet, 9-inches thick, except for the access tunnel part of the building which is 2-feet, 3-inches thick. The structure is supported on H-bearing piles. The access tunnel is separated from the rest of the Auxiliary Building by two inches of fiberglass expansion joint material. Intake Pumping Station The intake structure is supported by a reinforced concrete slab placed on a 4-inch-minimum-thick concrete subpour which caps the top of the irregular rock surface. The base slab is 4 feet thick with a 6-foot-wide by 10-foot deep key located at the back of the structure. This key extends the full width of the structure. The base slab extends 10 feet past the back wall and has two areas of 26 feet by 29 feet on each side that extend beyond the walls. The concrete retaining walls at the intake structure are designed to protect the forebay of the intake against earth slides during an earthquake. The base slabs of these cantilevered walls are keyed into rock. The walls are separated from the structure with expansion joint material. 3.8.5-2
WBN North Steam Valve Room The north steam valve room is supported by a grillage of reinforced concrete foundation walls to base rock. These walls span vertically from base rock at Elevation 683.0 to the bottom of the valve room base slab at Elevation 722.0. There are four 4-foot thick walls running in a north-south direction and these walls are tied together by a singular 4-foot thick wall running in an east-west direction. Three closed cells are formed by these walls in combination with the Reactor Building wall. These closed cells are backfilled with a non-compacted crushed stone. The valve room foundation walls are separated from the Reactor Building foundation and wall by a 2-inch fiberglass expansion joint material. Diesel Generator Building The base slab of the Diesel Generator Building is discussed in Section 3.8.5.5.2. Based on soils laboratory tests, it could not be assured that the existing material between the top of firm gravel at Elevation 713 and base slab was capable of safely supporting the structure. Therefore, this material was removed and replaced with crushed stone fill placed in 4-inch layers and compacted to a minimum of 70% relative density (see Section 2.5.4.5.2, historical information). A slope stability analysis was performed in order to assure stability of the slope below the building. Refueling Water Storage Tank The refueling water storage tank foundation is a solid, circular reinforced concrete structure placed on engineered granular fill over firm natural granular soil. The foundation is constructed with shear keys to prevent sliding displacement and with retaining walls to contain a reservoir of borated water after a postulated rupture of the storage tank. The foundation is protected from missiles by a concrete apron. Discharge Overflow Structure See Section 3.8.4.1.7 for a description of the discharge overflow structure foundation. Class 1E Electrical System Manholes and Duct Banks The manholes and a portion of the duct banks are supported on in-situ soil. The duct banks at the intake pumping station are supported on in-situ soil, piles, and a bracket on the pumping station wall, see Section 3.8.4.1.4 for additional information. ERCW Standpipe Structures See Section 3.8.4.1.7 for the standpipe structures. ERCW Pipe Supporting Slabs and Beams See Section 3.8.4.1.7 for a description of the beams and slab. 3.8.5-3
WBN ERCW Valve Covers See Section 3.8.4.1.7 for a description of these structures. Additional Diesel Generator Building The base slab of the additional Diesel Generator Building is discussed in Section 3.8.4.4.8. Similar to the Diesel Generator Building, it could not be assured that the existing soil between the top of firm gravel at Elevation 713.0 and the bottom of the base slab at Elevation 730.0 could safely support this structure. Therefore, the building was supported on end bearing steel H-Piles driven to refusal in sound rock or other suitable material. For additional information on this structure, see Section 3.8.4.1.8. 3.8.5.2 Applicable Codes, Standards, and Specifications See Sections 3.8.1.2, 3.8.3.2, and 3.8.4.2. 3.8.5.3 Loads and Loading Combinations The loads and loading combinations are described in Sections 3.8.1.3, 3.8.3.3, and 3.8.4.3. For loads and loading combinations on the Additional Diesel Generator Building, see Table 3.8.4-22. 3.8.5.4 Design and Analysis Procedure 3.8.5.4.1 Primary Containment Foundation The foundation was analyzed as a slab on a rigid foundation. The slab was analyzed using computer code Gendek 3 Finite Element Analysis of Stiffened Plates. Maximum tangential and radial moments were obtained using the finite element analysis of the various load combinations. Shear stresses were obtained by conventional analysis for the con-tainment vessel anchorage and major equipment loadings. 3.8.5.4.2 Auxiliary-Control Building The reinforced concrete base slab of the Auxiliary-Control Building was designed in compliance with the ACI Building Code 318-63. It was analyzed by the ICES STRUDL-II finite element method as a slab on an elastic foundation. In the ICES STRUDL-II program the foundation material was modeled by assigning a vertical spring to each node of the grid system which was used to represent the base slab. The base slab was divided into elements with wall stiffnesses being recognized by introducing flexural rigidity along the wall and torsional rigidity being recognized by including a rotational spring. Superposition of the various loading conditions were used to obtain maximum stresses. Manual calculations gave results for the bending moments which checked reasonably close with those obtained from the ICES STRUDL-II analysis. A standard frame analysis was also performed in order to determine the shearing forces in the slab. 3.8.5-4
WBN Shear walls fixed to the base slab transmit lateral force to the slab; the base slab itself is keyed and anchored into foundation rock to transmit shear from the structure into the rock. The 45-inch-thick slab of the waste packaging area was designed for a uniform distribution of base pressure to span as a flat plate between the load bearing walls. Walls were thicker than necessary for structural purposes because of shielding requirements. The base slab of the condensate demineralizer waste evaporator building portion was designed as a pile supported foundation. Batter piles were used around the perimeter of the structure to transmit lateral loads from the structure to the foundation media. 3.8.5.4.3 Intake Pumping Station The design of the base slab was controlled for the most part by uplift considerations under assumed unwatered conditions with one bay dry and full uplift over 100% of the area between the slab and the base rock. The backfilled portion of the base slab was controlled by the load from the saturated fill. 3.8.5.4.4 Soil-Supported Structures A uniform or linear distribution of base pressure was assumed in the design of all soil-supported structures and all base slabs were essentially designed as flat plates. 3.8.5.4.5 Pile Supported Structures Pile supported structures were designed using conventional frame analysis or through the use of ICES STRUDL-II finite element computer program. 3.8.5.5 Structural Acceptance Criteria 3.8.5.5.1 Primary Containment Foundation The base slab design contained the following conservative features:
- 1. No allowance was made for the additional spread of reactions under the walls or the additional section modulus due to the 3-foot structural fill over the base slab.
- 2. In the outer area of the slab, where the additional depth is in excess of the 2-foot, 8-inch recess in the upper surface, no allowance has been made for the additional thickness which increases the stiffness of the slab and thus lowers the stresses.
3.8.5-5
WBN 3.8.5.5.2 Foundations of Other Category I Structures Auxiliary-Control Building The base slab as designed has its maximum flexural stresses and shearing stresses within the allowable working stress design limits of Table 3.8.4-1 for all loading combinations. Design Case I (dead load plus live load), which generally controlled the design, was investigated by the ICES STRUDL-II program for several loading conditions created by the three different levels of the slab and by the early conditions were superimposed in various combinations to ensure that the slab was designed for the maximum possible stresses. The maximum calculated compression of the base slab was approximately 12 ksf. The maximum allowable compression on rock is 26 ksf (180 psi). In probable maximum flood conditions, with the dead load of the structure alone assumed to resist the buoyant force, the factor of safety against floatation is 1.51. Intake Pumping Station The base slab of the intake pumping station serves as a water barrier under maintenance conditions with one bay unwatered. It also adds to the stability of the structure. Backfill on the extended areas of the slab add weight to the structure and the key provides resistance to sliding. The maximum calculated compression on the base slab was approximately 12 ksf. The maximum allowable compression on rock is 26 ksf (180 psi). North Steam Valve Room The valve room foundation walls were designed to resist the maximum overturning effect on the building. This effect was due to pressure as the result of the rupture of a main steam pipe, its associated jet impingement load, and the Safe Shutdown Earthquake. This resistance to overturning was obtained by converting the maximum overturning moment on the structure into a resisting active soil pressure on the foundation walls. For overturning in the east-west direction, four of the foundation walls were considered effective. For overturning in the north-south direction, the singular cross-wall was considered to be resisting the overturning. Using this pressure as a load on the walls, they were modeled as plate structures utilizing the STRUDL-II Finite Element computer program. The walls were considered to span between bedrock, the bottom of the valve room base slab and other foundation walls framing into them. Waste Packaging Structure This structure is situated on well-compacted crushed stone backfill above rock and was designed for a normal allowable uniform bearing pressure of 6.5 ksf and a maximum allowable pressure with 70% or more of the base in compression of 10 ksf under maximum overturning forces. Actual calculated bearing pressures were 1.4 ksf for uniform loading and 6.7 ksf with 72% of the base in compression for maximum overturning forces. 3.8.5-6
WBN Diesel Generator Building The structure is situated as described in Section 3.8.5.1.2. The base slab of the Diesel Generator Building is 9 feet 9 inches thick founded on crushed stone backfill and located above the probable maximum flood elevation. The structure was designed for a normal allowable uniform bearing pressure of 6.5 ksf and a maximum allowable pressure of 11.5 ksf under maximum overturning forces. Actual calculated bearing pressures for the Diesel Generator Building were 2.0 ksf for uniform loading and 4.9 ksf for maximum overturning forces with 100% of the base in compression. Additional Diesel Generator Building For discussions on this pile supported structure see Section 3.8.4.4.8. Also, rotational restraint from the piles was not considered due to the large difference in stiffness between the 12 foot thickness of the base slab and that of the steel H-piles. 3.8.5.6 Materials, Quality Control, and Special Construction Techniques General See Section 3.8.1.6. 3.8.5.6.1 Materials Concrete and Reinforcing Steel See Section 3.8.1.6.1 Backfill Materials Backfill material was taken only from areas designated by the soils investigation program (see Section 2.5.4.5.2, historical information) as suitable for backfill material. 3.8.5.6.2 Quality Control Concrete and Reinforcing Steel Concrete production and testing were as in Section 3.8.1.6.2, except some concrete used to protect rock surfaces was purchased as ready mix in conformance with ASTM C94-69. 3.8.5-7
WBN The protective concrete for rock surfaces was specified as 2,000 psi at 90 days age. It was in conformance to specifications. The Shield Building base slab and the north steam valve rooms foundation walls used concrete specified as 5,000 psi at 90 days. Some concrete did not meet specification requirements. This was evaluated and documented in the Report CEB-86-19C "Concrete Quality Evaluation". Results have been documented in affected calculation packages and drawings. Testing of reinforcing steel was as in Section 3.8.1.6.2. Base Rock The base area of all rock-supported structures was inspected by the principal civil design engineer in conjunction with an experienced TVA geologist during final cleanup of rock surfaces to determine its suitability as a foundation. Backfill Quality control requirements for backfill material were as specified in Section 2.5.4.5 (historical information). 3.8.5.6.3 Special Construction Techniques No special construction techniques were used. REFERENCES None 3.8.5-8
WBN 3.8.6 Category I(L) Cranes 3.8.6.1 Polar Cranes 3.8.6.1.1 Description See Figures 3.8.6-1 through 3.8.6-6. There are two polar cranes, one in each of the Reactor Buildings. Each crane is a single two-part trolley, overhead, electric traveling type; operating on an 86-foot 0-inch-diameter rail at the top of the crane wall and above the reactor. Each crane has a main hoist capacity of 175 tons and an auxiliary hoist capacity of 35 tons. The Unit 1 polar crane main and auxiliary hoist motions are driven by dc motors with stepless regulated speed control. The bridge and trolley are driven by ac motors with static, stepless regulated speed control. The Unit 2 polar crane main and auxiliary hoist motions are driven by ac motors with Variable Frequency Drives. The bridge and trolley are driven by ac motors with variable Frequency Drives. Structural portions of the crane bridges consist of welded box-type girders and welded, haunched, box-type end ties. Structural portions of the trolleys consist of welded box-type trucks and welded cross beams. Control of each crane is from a cab located below the bridge walkway at one end of a girder. 3.8.6.1.2 Applicable Codes, Standards, and Specifications The following codes, standards, and specifications were used in the design of the cranes: National Electric Code, 1971 edition. National Electrical Manufacturers Association, Motor and Generator Standards, Standard MG-1, 1970 edition. Crane Manufacturers Association of American, Inc., Specification #70, 1970 edition. Federal Specification RR-W-410C American Society for Testing and Materials, 'Material Standards,' 1974 edition. American Welding Society, D1.1-72 with 1973 Revisions, Structural Welding Code. Section 1.23, Part 1, 'Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings,' Manual of Steel Construction, Part 5, American Institute for Steel Construction, 7th edition, 1970. American Gear Manufacturers Association Standards for Spur, Helical, Herringbone, and Beval Gears. 3.8.6-1
WBN Where date of edition, copyright, or addendum is specified, earlier versions of the listed documents were not used. In some instances, later revisions of the listed documents were used where design safety was not compromised. The cranes meet applicable requirements of the listed codes, standards, and specifications. 3.8.6.1.3 Loads, Loading Combinations, and Allowable Stresses Loads, loading combinations, and allowable stresses are shown in Table 3.8.6-1. 3.8.6.1.4 Design and Analysis Procedure The bridge girders and end ties for each crane were designed as simple beams in the vertical plane and as a continuous frame in the horizontal plane. Stresses in the girders and end ties were computed with the trolley positioned to produce maximum stresses. Seismic restraints are located on the bottom of each girder and these restraints are designed to withstand seismically applied loads to ensure the crane will not fall during an earthquake. Trolley positions used were the maximum end position, third point, and the point near the center which produces maximum bending moments. Trolley members were designed as simple beams. Design of the bridge girders and end ties was by TVA. Mechanical parts and structural members except the bridge girders and end ties were designed by the contractor. Calculations and designs made by the contractor were reviewed by TVA design engineers. In designing for earthquake conditions, forces due to accelerations at the crane bridge rails were used as static loads for determining component and member sizes. After establishing component and member sizes, a dynamic analysis, using appropriate response spectra, was made of the total crane to determine that allowable stresses had not been exceeded. Earthquake accelerations at the bridge rails were determined by dynamic analysis of the structures supporting the crane rails. The polar crane was also evaluated for seismic loads based on the Set B seismic response spectra using 2% damping for OBE and 4% damping for SSE. The polar crane was initially evaluated for seismic loads based upon Set A seismic response spectra. 3.8.6-2
WBN 3.8.6.1.5 Structural Acceptance Criteria Allowable stresses for all load combinations used for the various crane parts are given in Table 3.8.6-1. For normal load conditions, the allowable stresses provide safety factors of 2 to 1 on yield for structural parts and 5 to 1 on ultimate for mechanical parts, except for wire ropes which have a minimum safety factor of 5 to 1 on ultimate. For limiting conditions, such as an SSE earthquake or stall, stresses do not exceed .9 yield. 3.8.6.1.6 Materials, Quality Controls, and Special Construction Techniques A36 steel was used for the major structural portions of the crane. Design by TVA and erection by TVA were in accordance with TVA's quality assurance program. Design and fabrication by the contractor were in accordance with the contractor's quality assurance program which was reviewed and approved by TVA's design engineers. The contractor's quality assurance program covers the criteria in Appendix B of 10 CFR 50. Fabrication procedures such as welding, stress relieving, and nondestructive testing were included in appendices to the contractor's quality assurance program. ASTM standards were used for all material specifications and certified mill test reports were provided by the contractor for materials used for all load-carrying members. This crane is covered by TVA's Augmented Quality Assurance Program for Seismic Category I(L) Structures. 3.8.6.1.7 Testing and In-service Surveillance Requirements Refer to Section 14.2.7 (historical information), Paragraph 4.A.1.h for Initial Testing. After the initial test, periodic visual inspections of each crane are to be made. Parts inspected during the visual inspection are to include all bolted parts, couplings, brakes, hoist ropes, hoist blocks, limit switches, and equalizer systems. 3.8.6.1.8 Safety Features The cranes were designed to withstand an SSE and to maintain any load up to rated capacity during and after the earthquake period. 3.8.6-3
WBN-1 The bridges are equipped with double flange wheels, spring-set, electrically-released brakes which set and firmly lock two of the wheels when the bridge drive machinery is not operating or when power is lost for any reason, hold down lugs which run under the rail heads, and seismic restraints located on the bottom of each girder. During an earthquake the crane rail will yield before the crane wheels fail, thus allowing the crane to move until the seismic restraints on each girder contact the crane wall. These restraints hold the crane on the runway. Guide rollers, mounted on each extreme corner truck, travel against the outer surface of the bridge rail to assure bridge truck alignment. The trolleys are each equipped with double flange wheels, spring-set, electrically-released brakes which set and firmly lock the driving wheels when the trolley drive machinery is not operating or when power is lost for any reason, and hold down lugs which run under the rail heads. Positive wheel and bumper stops are provided at both ends of the bridge. During an earthquake, the trolley could be displaced, but it will not leave its rails which are firmly attached to the bridge structure. Safety features provided for each hoist include two independent gearing systems, connected by a cross shaft to prevent windup, two brakes with each of the brakes operating through one of the independent gearing systems, two upper travel limit switches, one lower travel limit switch, over-speed switches set to trip at 120% of maximum rated speed. The Unit 1 polar crane has emergency dynamic braking for controlled lowering in case of simultaneous failure of ac power source and holding brakes. For the Unit 2 polar crane, upon loss of AC power, load is manually lowered by manipulating the holding brakes. In addition, each hoist incorporates a symmetrical cross reeving system designed to hold the load level with either rope. Each hoist is also provided with a hydraulic equalizing system to prevent dropping the load and to limit shock loading in case of a single rope failure. Holding brakes for the hoists are the spring-set, electrically released type with provisions for manual release of the brakes. The capacity of each main hoist brake is sufficient to stop a 100% rated load traveling at the maximum rated hoisting speed within a distance of 6 inches. Safety control features provided for all motions consist of overcurrent protection, undervoltage protection, control actuators which return to the stop position when released, and an emergency-stop push button. 3.8.6.2 Auxiliary Building Crane 3.8.6.2.1 Description The crane in the Auxiliary Building is a single trolley, overhead, electric traveling type with a span of 77 feet. The crane has a main hoist capacity of 125 tons and an auxiliary hoist capacity of 10 tons. The main hoist meets NUREG-0554 single failure proof criteria for compliance with 10 CR 72.124(a), Design for Criticality Safety for handling spent fuel casks. 3.8.6-4
WBN-1 The main and auxiliary hoists are driven by AC Variable Frequency motor drives with eddy current lowering and stepless (infinite) speed control. The bridge and trolley travel motions are AC operated with stepless control. Structural portions of the crane bridge consist of welded, box-type girders and welded, haunched, box-type end ties. Structural portions of the trolley consist of welded, box-type trucks and welded cross beams. Control of the crane is from a control console in the operator cab which is located at mid-span of the crane beneath the south girder. The one crane serves the needs of two reactor units. It handles the fuel casks, new fuel shipments to the new fuel storage, shield plugs at the equipment access doors, and any large pieces of equipment going into or out of the Reactor Buildings via the Auxiliary Building. 3.8.6.2.2 Applicable Codes, Standards, and Specifications The following codes, standards, and specifications were used in the original design of the crane: National Electric Code, 1971 Edition. NEMA Standard MG1, 1970 Edition. Crane Manufacturers Association of American, Inc., Specification No. 70, 1970 Edition. Federal Specification RR-W-410C. Auxiliary (10 Ton) Hoist Only ASTM Material Standards, 1974 Edition. AWS, D1.1-72 with 1973 Revisions, Structural Welding Code. Section 1.23, Part I, Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings; AISC Manual of Steel Construction, 7th Edition, 1970. American Gear Manufacturers Association Standards for Spur, Helical, Herringbone, and Bevel Gears. DIN EN (10264-2/12385-2), Steel Wire and Wire Products- Steel Wire for Ropes, Part 2/Steel Wire Ropes - Safety - Part 2 (Main Hoist Wire Rope only) The following codes and Standards were used to qualify the 125 Ton Auxiliary Building Crane Main Hoist as NUREG-0554 single-failure-proof: CMAA #70 (2010) NUREG-0554 NUREG-0612 3.8.6-5
WBN-1 Where date of edition, copyright, or addendum is specified, earlier versions of the listed documents were not used. In some instances, later revisions of the listed documents were used where design safety was not compromised. The cranes meet applicable requirements of the listed codes, standards, and specifications. 3.8.6.2.3 Loads, Loading Combinations, and Allowable Stresses Loads, loading combinations, and allowable stresses are shown in Table 3.8.6-2. 3.8.6.2.4 Design and Analysis Procedure The 125 Ton Auxiliary Building Bridge Crane is designed to retain control of and hold the load, and the bridge and trolley are designed to remain in place on their respective runways with their wheels prevented from leaving the tracks during a seismic event. The Maximum Critical Load (125 Tons) plus operational and seismically induced pendulum and swinging load effects on the crane were considered in the design of the trolley, and they were added to the trolley weight for the design of the bridge. The crane was analyzed with the trolley located at mid-span, one-third span, close approach, and with the hook in the high and low position at each location. These analyses were competed using 4% acceleration response spectra which were conservatively derived using the average of the 3% and 5% response spectra. It was determined from this analysis that all structural components analyzed are capable of withstanding the seismically induced loadings. The Auxiliary Building 125/10 Ton Crane Main Hoist is designed to NUREG-0554 single-failure-proof requirements. The head and load block, reeving, and hook are designed to support a static load of 200% of the Maximum Critical Load (MCL) of 125 Tons. The main hoist drive train, bridge, and trolley components have a 15% margin relative to the maximum critical load. The auxiliary hoist is designed to CMAA #70 requirements. 3.8.6-6
WBN-1 3.8.6.2.5 Structural Acceptance Criteria Calculated stresses for the crane are all within CMAA #70 (2010) limits. Additionally, the individual component parts of the vertical hoisting system components, which include the head block, rope reeving system, load block, and sister hook, are designed to support a static load of 200% of the Maximum Critical Load of 125 Tons for the main hook. Drive train components for the main hoist, bridge drive and trolley drive, including requirements to deal with degradation due to wear and exposure. 3.8.6.2.6 Materials, Quality Controls, and Special Construction Techniques ASTM A 36 steel was used for the major structural portions of the crane. Design by TVA and erection by TVA were in accordance with the TVA quality assurance program. Design and fabrication by the contractor were in accordance with the contractor's quality assurance program which was reviewed and approved by TVA's design engineers. The contractor quality assurance program covers the criteria in Appendix B of 10 CFR 50. Fabrication procedures such as welding, stress relieving, and nondestructive testing, were included in appendices to the contractor's quality assurance program. ASTM standards were used for all material specifications and certified mill tests reports were provided by the contractor for materials used for all load-carrying members. This crane is covered by TVA's Augmented Quality Assurance Program for Seismic Category I(L) Structures. 3.8.6.2.7 Testing and In-service Surveillance Requirements Upon completion of erection and adjustments on the crane, all crane motions and operating parts were thoroughly tested with crane handling 125% of rated capacity. Tests were made to prove the ability of the crane to handle its rated capacity and smaller loads smoothly at any speed within the specified speed range. Each brake was tested to demonstrate its ability to hold the required load. After the initial test, periodic visual inspections of the crane are to be made. Parts inspected during the visual inspection are to include all bolted parts, couplings, brakes, hoist ropes, hoist blocks, limit switches, and equalizer systems. 3.8.6-7
WBN-1 3.8.6.2.8 Safety Features The crane was designed to withstand an SSE and to maintain any load up to rated capacity during and after the earthquake period. The bridge is equipped with double flange wheels, hold down lugs which run under the rail heads two spring-set electrically released brakes which set and firmly lock the wheels when the bridge drive machinery is not operating or when power is lost for any reason. During an earthquake the crane rail will yield before failure of the crane wheels and allow the end ties to contact the adjacent concrete wall, thus restraining the crane and preventing it from falling. Positive wheel and bumper stops are provided at each end of the bridge travel. The trolley is equipped with double flange wheels, two spring-set, electrically released brakes which set and firmly lock the driving wheels when the trolley drive machinery is not operating or when power is lost for any reason, and hold down lugs which run under the rail heads. Positive wheel and bumper stops are provided at both ends of the bridge. During an earthquake, the trolley could be displaced, but it will not leave the rails which are firmly attached to the bridge structure. Safety features provided for each hoist include two independent gearing systems, connected by a cross shaft to prevent windup, two brakes with each of the brakes operating through one of the independent gearing systems, two upper traveling limit switches, one lower travel limit switch, over-speed switches set to trip at 120% of maximum rated speed, and emergency dynamic braking for controlled lowering in case of simultaneous failure of ac power source and holding brakes. In addition, the main hoist incorporates a symmetrical cross reeving system designed to hold the load level with either rope and to limit the shock loading in case of a single rope failure, and a hydraulic sheave equalizing system to prevent dropping the load and to limit shock loading in case of a single rope failure. The auxiliary hoist has a two-part whip-style reeving so that a single rope failure will not drop the load. Holding brakes for the hoists are the spring-set, electrically released type with provisions for manual release of the brakes. The capacity of each main hoist brake is sufficient to stop at 100% rated load traveling at the maximum rated hoisting speed within a distance of 6 inches. The interlocks will not be bypassed for any heavy loads except the fuel transfer gates and for new fuel handling. All loads in excess of 2,059 lbs, or which would have a kinetic energy greater than that of a spent fuel assembly from its normal handling height, will be transported around the spent fuel pit, rather than over, with the interlocks activated, via the normal paths used for heavy loads. The Main Hoist of the 125/10 Ton Auxiliary Building Crane meets NUREG-0554 requirements as single-failure-proof. 3.8.6-8
WBN Safety control features provided for all motions consist of overcurrent protection, undervoltage protection, control actuators which return to the stop position when released, and an emergency-stop pushbutton. The electrical interlocks and mechanical stops will be administratively bypassed to allow use of the crane for handling the fuel transfer canal gate. The bypass is accomplished by means of a keyed switch, operation of which bypasses all interlocks controlling crane movements and activates a green indicating light located beneath the operator's cab. The indicating light is visible from any point on the operating floor. Control of the bypass key by administrative personnel and the ability of administrative personnel to stop the crane by means of any one of three pushbutton stations ensure that administrative personnel control all bypass operations. Two pushbutton stations are located on the west wall and one pushbutton station is located on the east wall of the Auxiliary Building about four feet above the Elevation 757.0 operating floor. These stations are readily accessible to administrative personnel on the operating floor. Testing of bypass interlocks is accomplished on a periodic basis in accordance with approved WBNP surveillance instructions. Testing must occur within seven calendar days prior to initial use, and every seven calendar days during continued regular usage. Each limit switch is manually operated to ascertain proper functioning of interlock circuits. To verify that the interlock system is functioning properly, each limit switch is moved to its actuated position, and all affected crane controls operated to ensure that crane movement does not occur. REFERENCES None 3.8.6-9
WBN TABLE 3.8.6-1 (Sheet 1 of 3) POLAR CRANES LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES No. Load Combinations Allowable Stresses (psi) Tension Compression(1) Shear Bridge Structure I Dead Live 0.50 Fy 0.48 Fy 0.33 Fy Impact Trolley tractive II Dead Live 0.50 Fy 0.48 Fy 0.33 Fy Impact Bridge tractive III Dead Live 0.62 Fy 0.59 Fy 0.41 Fy Trolley collision IV Dead Trolley weight 0.90 Fy 0.90 Fy 0.50 Fy Stall at 275% capacity V Dead Live at 100% capacity 0.90 Fy 0.90 Fy 0.50 Fy SSE
WBN TABLE 3.8.6-1 (Sheet 2 of 3) POLAR CRANES LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES (Cont'd) No. Load Combinations Allowable Stresses (psi) Tension Compression(1) Shear Trolley Structure I Dead Live 0.5 FY 0.48 FY 0.33 FY Impact II Dead 0.9 FY(3) 0.9 FY 0.5 FY Stall at 275% capacity 0.62 FY (2) 0.59 FY 0.41 FY III Same as case V for bridge Mechanical Parts No. Load Combinations Allowable Stresses (psi) Tension and Compression(1) Shear Parts Other Than Wheel Axles and Saddle Truck Connecting Pins I Dead Ult 2 x Ult Live 5 15 II Dead 0.9 FY 0.5 FY Stall at 275% capacity Wheel Axles and Connecting Pins I Dead Ult 2 x Ult Live 5 15 Impact II Dead Ult 2 x Ult Live 5 15 Collision
WBN TABLE 3.8.6-1 (Sheet 3 of 3) POLAR CRANES LOADS, LOADING COMBINATIONS, AND ALLOWABLE STRESSES (Cont'd) No. Load Combinations Allowable Stresses (psi) Tension and Compression(1) Shear Wheel Axles and Connecting Pins (Continued) III Dead 0.40 FY 0.50 FY Stall at 275% capacity IV Dead 0.90 FY 0.50 FY Live at 100% capacity SSE Notes: (1) The value given for allowable compression stress is the maximum value permitted, when buckling does not control. The critical buckling stress, Fcr, shall be used in place of FY when buckling controls. (2) For sheave frames, cross beams, and their respective connections. (3) For all other members.
WBN-1 TABLE 3.8.6-2 DELETED
WBN APPENDIX 3.8A SHELL TEMPERATURE TRANSIENTS Figure 3.8A-1 presents average shell temperatures adjacent to the three compartments as a function of time after the DBA. The DBA is a double end rupture of the reactor coolant pipe with the reactor decay heat released into the lower compartment as steam. Initially the steam is condensed in the ice compartment. After the ice melts, the steam is condensed in the upper compartment by a water spray. The lower compartment temperature rises to 250°F, essentially instantaneously, then is reduced to 220°F very shortly after the blowdown is completed. The blowdown is completed before the shell adjacent to the lower compartment reaches 220°F, as illustrated by the smooth curve presented in Figure 3.8A-1. The upper compartment temperature rises essentially instantaneously due to compression of the noncondensable gases into the upper compartment. The sharp rise at 7,000 seconds simulates the disappearance of the ice from the ice compartment. The shell temperature will rise at a maximum of 0.11 degree per second during the rise from 140°F to 190°F. The subsequent temperature decrease of the shell adjacent to the upper compartment is due to the reduction in decay heat. The curve labeled shell adjacent to the ice compartment indicates the temperature of the shell adjacent to the ice compartment. The shell is separated from the ice compartment with a thick layer of insulation, hence the rather slow response for the temperature of the shell adjacent to the ice compartment. After the ice is all melted the temperature inside the ice compartment will be the same as the temperature in the lower compartment; however, the shell temperature adjacent to the ice compartment will always be less than the temperature in the ice compartment because of insulation. The temperature of the shell adjacent to the ice compartment will peak at less than 220°F. The curves in Figure 3.8A-1 are an average shell temperature representative for the bulk of the shell. Some areas near boundaries between compartments and near the base will differ significantly from the bulk. The lower portion of the lower compartment shell will be insulated for the purpose of minimizing the transient effects. Figure 3.8A-2 is a plot of shell temperature versus distance above Elevation 702.78 for various times after a LOCA. In establishing these curves it was assumed that top of the concrete slab is at Elevation 702.78 inches, and that the top of the insulation is at Elevation 707.11, and the top 8 inches of insulation is tapered from 2 inches thick to 1/4-inch thick. 3.8A-1
t 300 SHELL WALL TEHPE.~TURE VERSUS TIHE AFTER LOSS OF COOLANT SHELL ADJACENT TO LO./ER co:*IPAP.THENT C"" z SHELL ADJACENT TO UPPER COMPARTMENT 100 I 1------------------- so SHELL ADJACENT TO ICE CONDENSER 0 ~ - - - - - - - - _ . __ _ _ _ _ _ _ __.__ _ _ _ _ _ _ _ _..J..._ _ _ _ _ _ _ ___.__ _ _ _ __ 0 10 100 1000 10,000 TIME (SEC) WATTS BAR NUCLEAR PLANT FINAL SAFETY WATTS BAR ANALYSIS NUCLEAR PLANT REPORT FINAL SAFETY ANAL vs,~ REPORT Shell SHELL WALL TEMPERATURE Wall Temperature Versus VERSUS Time After Loss of Coolant TIMI AFTER LOSS OF COOLANT l'igure 3.BA-l FIGURE 3.8A-1
TYPICAL TEMPERATURE TRANSIENT LOWER COMPARTMENT WALL 5 MINUTES STEADY STATE 1 MINUTE 5
,STOP OF INSULATION W
d' O 4 5 MINUTES J N J 30 MINUTES z z 3 2 HOURS w O m W U z Q H O ANNULUS FLOOR EL, 702'-9 3/8" 70 80 100 C 120 140 160 180 200 220 TEMPERATURE IN 'F WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Typical Temperature Transient Lower Compartment FIGURE 3.8A-2
WBN APPENDIX 3.8B BUCKLING STRESS CRITERIA 3.8B.1 INTRODUCTION The buckling design criteria in this appendix are applicable to stiffened circular cylindrical and spherical shells. Section 2.0 sets forth the buckling design criteria for shells stiffened with circumferential stiffeners. Because of existing penetrations, interferences, or large attached masses, it may be expedient to further analyze some areas of the vessel as independent panels. Section 3.0 sets forth the criteria for shells stiffened with a combination of circumferential and vertical stiffeners. Section 4.0 deals with the criteria for a spherical dome. The procedures and data presented were adapted primarily from Chapter 3 of the Shell Analysis Manual, by E. H. Baker, A. P. Cappelli, L. Kovalevsky, F. L. Rish, and R. M. Verette, National Aeronautics and Space Administration, Washington, D.C., Contractor Report CR-912, April 1968. The criteria given in this section cover only the range of variables needed for the structural steel containment vessel for which these specifications were prepared. The buckling criteria are specified in terms of unit stresses and membrane forces in the shell. Stresses caused by multiple loads must be combined according to provisions of Table 3.8B-1 for use in these criteria. The values of the load factors and factors of safety used in the buckling criteria are given in Section 5.0. The method of applying the factors of safety to the criteria is shown in Table 3.8B-2. 3.8B.2 SHELLS STIFFENED WITH CIRCUMFERENTIAL STIFFENERS 3.8B.2.1 Circular Cylindrical Shells Under Axial Compression The critical buckling stress for a cylinder under axial compression alone is determined by the equation Cc Et (1) cr = R for various ranges of cylinder length defined by L2 1-2 Z= Rt The constant Cc is determined from Figure 3.8B-l for the appropriate value of R/t. The critical buckling stress in a cylinder under axial compression and internal pressure is determined by: Et (1) cr = (Cc + Cc) R 3.8B-1
WBN The constants Cc and Cc are determined from Figures 3.8B-1 and 3.8B-2, respectively. The constant Cc given in Figure 3.8B-2 depends only upon the internal pressure and R/Et. 3.8B.2.2 Circular Cylindrical Shells in Circumferential Compression A circular cylindrical shell under a critical external radial or hydrostatic pressure will buckle in circumferential compression. The critical circumferential compressive stress is given by: 2 2 ( 2) Kp E t cr = 12(1 - u 2) L for various values of Z given in Section 2.1. Curves for determining the constant Kp for both radial and hydrostatic pressure are given in Figure 3.8B-3. 3.8B.2.3 Circular Cylindrical Shells Under Torsion The shear buckling stress of the cylinder subject to torsional loads is given by: Et (3) cr = Cs RZ1/4 The shear buckling stress of the cylinder subject to torsion and internal pressure is determined by Et cr = (Cs + Cs) Cs (3) 1/ 4 RZ where constants, Cs and Cs, are determined from Figures 3.8B-4 and 3.8B-5. Values of Cs are given for internal radial pressure alone and internal pressure plus an external load equal to the longitudinal force produced by the internal pressure. Figure 3.8B-4 is applicable for values of: L2 1-2 Z = > 100 Rt 3.8B-2
WBN For cylinders with length constant Z less than 100, the shear buckling stress is determined by: 2 K s 2 E t (3) cr = a b 12 (1 - 2) a for values of: a2 1-2 Z = Rt where a is the effective length and b is the circumference of the cylinder. The coefficient Ks' is given in Figure 3.8B-10. 3.8B.2.4 Circular Cylindrical Shells Under Bending The critical buckling stress for the cylinder under bending is computed by the equation: Et (4) cr = C b R where the buckling constant, Cb is given by Figure 3.8B-6. The critical buckling stress for the cylinder under internal pressure and bending is computed by: Et (4) cr = ( C b + C b ) R where Cb and Cb are given by Figures 3.8B-6 and 3.8B-7, respectively. Figure 3.8B-7 is a function of the internal pressure and the geometry. 3.8B.2.5 Circular Cylindrical Shell Under Combined Loads The criterion for buckling failure of the cylindrical shell under combined loading is expressed by an interaction equation of stress-ratios of the form: R1x + R2y + R3z 1 Note that m (n) F m k (n) F k (n) (n) N1 F1 N2 F2 Rn = + + . . . . . . cr (n) t cr (n) t (n) (n) cr t cr t 3.8B-3
WBN where Nm is the compressive or shear membrane force and Fm is the appropriate load factor, given in Section 5.0, for individual loading components in any loading combination. The superscript n refers to the particular type of loading. Superscripts n = 1, 2, 3, and 4 represent respectively axial compression, circumferential compression, torsion, and bending loads. The following interaction equations were used in the design of the cylindrical shell.
- a. Axial Compression and Circumferential Compression m=k m=k N (1) m Fm N (2) m Fm m=o cr t (1) +
m=o cr t (2) < 1
- b. Axial Compression and Bending m= k (1) m=k ( 4)
Nm Fm Nm Fm (1) m = o cr t
+ ( 4) m = o cr t
< 19
- c. Axial Compression and Torsion 2
m=k N (1) m Fm m=k N (3)m Fm m=o cr t (1) + (3) m=o cr t
< 1
- d. Axial Compression, Bending, and Torsion 2
m=k N (1) m Fm m=k N (4) m Fm m=k N (3) m Fm m=o cr t (1) + m=o cr t (4) + (3) m=o cr t
< 1
- e. Axial Compression, Circumferential Compression, and Torsion 2
m=k N (1) m Fm m=k N (2) m Fm m=k N (3) m Fm m=o cr t (1) + m=o cr t (2) + (3) m=o cr t
< 1 The longitudinal membrane stresses produced by the nonaxisymmetric pressure loads (NASPL) were considered as caused by bending loads in the interaction equations.
3.8B-4
WBN 3.8B.3 SHELLS STIFFENED WITH A COMBINATION OF CIRCUMFERENTIAL AND VERTICAL STIFFENERS 3.8B.3.1 The shell was provided with permanent circumferential and vertical stiffeners. The circumferential stiffeners were designed to have a spring stiffness at least great enough to enforce nodes in the vertical stiffeners so as to preclude a general instability mode of buckling failure, thus ensuring that if buckling occurs, it will occur in stiffened panels between the circumferential stiffeners. An acceptable procedure for determining the critical buckling stresses in the vertical stiffeners and stiffened panels is outlined in Section 3.43 Shell Analysis Manual, by E. H. Baker A. P. Cappelli, L. Kovalevsky, F. L. Rish, and R. M. Verette, National Aeronautics and Space Administration, Washington, D.C., Contractor Report CR-912, April 1968. 3.8B.3.2 In addition for shells stiffened with a combination of circumferential and vertical stiffeners under combined load, the criterion for buckling failure of the shell plate is expressed by an interaction equation of stress ratios in the form similar to the interaction equations of Section 2.5. R1x + R 2y + R 3z < 1 The critical buckling stresses for the shell plates between the circumferential and vertical stiffeners were determined by the following equations.
- a. Curved Panel under Axial Compression.
The critical buckling stress for a curved cylindrical panel under axial compression alone is determined by the equation: 2 Kc 2 E t cr = 12(1- ) 2 b for various ranges of cylinder length given by: b2 1-2 Z = - Rt IVF-- The constant Kc is determined from Figure 3.8B-8.
- b. Curved Panel in Circumferential Compression The critical buckling stress of a curved cylindrical panel under circumferential compression was determined by Section 2.2.
3.8B-5
WBN
- c. Curved Panel Under Torsion The shear buckling stress of a curved cylindrical panel subjected to torsional loads is given by:
2 Ks 2 E t cr = a b 12 (1 - ) 2 b for values of: b2 1 - 2 Z = Rt The coefficient, Ks, is given in Figure 3.8B-9. For cylindrical panels with length, a, less than the arc length, b, the shear buckling stress is determined by: 2 2 Ks E t cr = ab 12 (1 - 2) a for values of: a2 1-2 z = Rt Curves for determining Ks are given in Figure 3.8B-10.
- d. Curved Panels Under Bending The critical buckling stress for a curved panel in bending was computed using the equation for axial compression given in (a) of this section.
3.8B.3.3 The critical buckling stress in a stiffened hemispherical shell for the analysis was not treated in the Shell Analysis Manual, and except for external pressure, was determined by the following equation: t cr = 0.125 E R where t = thickness of shell E = modulus of elasticity R = radius of shell 3.8B-6
WBN 3.8B.4 SPHERICAL SHELLS 3.8B.4.1 The critical buckling stress in the spherical dome, except for external pressure, was determined by the following equation: Et cr = 0.125 R where t = thickness of shell E = modulus of elasticity R = radius of shell 3.8B.4.2 Spherical Shell Under Combined Loads The criterion for buckling failure of the dome is expressed by an interaction equation of the stress ratios in the form: R1x + R 2y + R 3z < 1 similar to the interaction equation of Section 2.5. A set of interaction equations similar to those in Section 2.5 was used in the design except that the effects due to torsion were considered. 3.8B.5 FACTOR OF SAFETY The buckling stress criteria were evaluated to determine the factors of safety against buckling inherent in the criteria. The factors which affect stability were determined and the criteria were evaluated to account for these factors. The basis used to evaluate the criteria to account for the factors were (1) how well established are the effects of the factors on stability of these shells (2) amount of supporting data in the literature and (3) margins marked by the critical stresses and interaction equations used in the criteria. The buckling criteria were found to be very conservative and judged to provide at least a factor of safety of 2.0 against buckling for all loading conditions for which the vessels were designed. In addition, a load factor of 1.1 was applied to load conditions which include the Safe Shutdown Earthquake (SSE). A load factor of 1.25 was used with all other load conditions. 3.8B-7
WBN TABLE 3.8.B-1 (Sheet 1 of 4) MULTIPLE LOAD COMBINATIONS VARIOUS PLANT CONDITIONS LOADING CONDITIONS MSLB MSLB Norm. Accident Accident Post Load Const. Test. Normal Oper. 1/2 Accident (Static) Accident (Static) Accident Components Cond. Cond. Design SSE 1/2 SSE 1/2 SSE SSE SSE Flooding Personnel Access X X Lock Load Penetration Loads X X X X X X Containment X Vessel and X X X X X X X X Appurtenances Dead Loads Walkway Live X Loads Spray Header and Lighting Fixtures X X X Live Loads Safe Shutdown Earthquake (SSE) X X Lateral and Vertical Loads Design Internal Pressure or Design X X External Pressure One-half Safe Shutdown Earthquake (1/2 X X X X SSE) Lateral and Vertical Loads
WBN TABLE 3.8.B-1 (Sheet 2 of 4) MULTIPLE LOAD COMBINATIONS VARIOUS PLANT CONDITIONS LOADING CONDITIONS MSLB MSLB Norm. Accident Accident Post Load Const. Test. Normal Oper. 1/2 Accident (Static) Accident (Static) Accident Components Cond. Cond. Design SSE 1/2 SSE 1/2 SSE SSE SSE Flooding Design Internal Pressure or X X Pressure Transient Loads Design X Temperature Internal X X Temperature Range of 60°F to (for (for X 120°F Pressure Pressure Transient Transient Loads) Loads) Internal X X Temperature (for (for Range of 80°F to Design Design 250°F Internal Internal Pressure) Pressure) Thermal Stress Loads Including X X Shell Temperature Transients Hydrostatic Load Case 1A or 1B X X (See Note 1)
WBN TABLE 3.8.B-1 (Sheet 3 of 4) MULTIPLE LOAD COMBINATIONS VARIOUS PLANT CONDITIONS LOADING CONDITIONS MSLB MSLB Norm. Accident Accident Post Const. Test. Normal Oper. 1/2 Accident (Static) Accident (Static) Accident Load Components Cond. Cond. Design SSE 1/2 SSE 1/2 SSE SSE SSE Flooding Hydrostatic Load Case II (See Note X 1) Wind Loads X (See Note 2) Snow Loads X (See Note 2) Temporary Construction X Loads Internal Test X Pressure Weight of X Contained Air MSLB Pressure X X Internal Temperature X X Range of 80°F to 327°F Airlock Live Load X X Ice Condenser X X Duct Loads
WBN TABLE 3.8.B-1 (Sheet 4 of 4) MULTIPLE LOAD COMBINATIONS VARIOUS PLANT CONDITIONS Notes:
- 1. Hydrostatic loads case 1A & 1B, and load case II are shown on TVA drawing 48N400.
- 2. Wind and snow loads do not act simultaneously.
- 3. For allowable stress condition see Table 3.8.B-2.
WBN TABLE 3.8B-2 ALLOWABLE STRESS INTENSITIES PLUS BUCKLING LOAD FACTORS Applicable ASME Code Reference (1) for Stress Buckling Load Loading Conditions Intensity Factors Normal Design Condition NB-3221 In accordance with Construction Condition ASME Code, Section VIII Normal Operation Condition NB-3222 Load factor = 1.25 for both cylindrical portion and hemispherical head Upset Operation Condition NB-3223 Load factor = 1.25 for both cylindrical portion and hemispherical head Emergency Operation NB-3224 Load factor = 1.10 for both Condition cylindrical portion and hemispherical head Test Condition NB-3226 NA Post-Accident Fuel Recovery NB-3224 NA Condition (1) All code references are to the ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, 1971 Edition, with Winter 1971 Addenda.
BUCKLING-STRESS COEFFICIENT, CC, FOR UNSTIFFENED UNPRESSURIZED CIRCULAR CYLINDERS SUBJECTED TO AXIAL COMPRESSION WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Buckling-Stress Coefficient FIGURE 3.8B-1
INCREASE IN AXIAL-COMPRESSIVE BUCKLING-STRESS COEFFICIENT OF CYLINDERS DUE TO INTERNAL PRESSURE 10 a 6 4 1'~OR I NMI
~ 1 NMI
- r. ~NEI
.. X111 mill ? 4 6 8 I 4 6* 10 10? WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Axial-Compressive Buckling Stress FIGURE 3.8B-2
BUCKLING COEFFICIENTS FOR CIRCULAR CYLINDERS SUBJECTED TO EXTERNAL PRESSURE nT WZE IP 102 e At wer K2 t0 P -M.WPO.""d~
, lAl Z s 1 s i F Z a 4 a j 2 4
- 4 6
1* j *, 6 10; 10 10 Z WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Buckling Coefficients for Circular Cylinders FIGURE 3.8B-3
BUCKLING-STRESS COEFFICIENT, CS FOR UNSTIFFENED UNPRESSUREIZED CIRCULAR CYLINDERS SUBJECTED TO TORSION Ca 0.10 0.50 {,: , :~ .~. **: T. rev 0.56 R 0.54 j
.., Volga for ZN00 No $AwVlr s.~rta Eave+
,.* *. *.. ZMOO flee C1..p1 tag" 0.5? :::::':
0.50
.* .1. :*. 111. .1 1 ... .** is ... *. *. :3~ . .. 1 L.F.
~: : ;:
- *.~
.1.
1' : is :: 0.10
- ... *: *~ :' ~:' :' 1 ::i:* :iii .. , 1.:. ..:.
I ,r 0." ~~ }
.. .... '1 0.44 1000 ?000 7000 1000 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Buckling-Stress Coefficients FIGURE 3.8B-4
INCREASE IN TORSIONAL BUCKLING-STRESS COEFFICIENT OF CYLINDERS DUE TO INTERNAL PRESSURE 10 8 =2111l mass I ==Manauu vie 6 gig 4 oil r, 1.0 e 6 4 oil C, 0.10 2 MEN I 0.or 2 4 6 8 - 2 4 6 8 Z s 6 S T 4 o a 0.01 0.10 t.o 10 102 (!_)2 E WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Torsional Buckling-Stress Coefficients FIGURE 3.8B-5
BUCKLING-STRESS COEFFICIENT, Cb, FOR UNSTIFFENED UNPRESSURIZED CIRCULAR CYLINDERS SUBJECTED TO BENDING am o rw e y FM q 01~
`w Ip e g o M e Q o g ME Murm O
0 I I WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Buckling-Stress Coefficient, Cb FIGURE 3.8B-6
INCREASE IN BENDING BUCKLING-STRESS COEFFICIENT OF CYLINDERS DUE TO INTERNAL PRESSURE 10 t d 4 ej mm gill 2
~8832~ ~3E 2
0.10 t 6 EM liI Nil 4 min 0.01 2 0.01 7 4 6 t 0.10 7 4 6 t t0 2 4 ii 6 9 t0 2 ii 4 6 8 102 E~R`2 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Bending Buckling-Stress Coefficient FIGURE 3.8B-7
BUCKLING-STRESS COEFFICIENT, KC, FOR UNPRESSURIZED CURVED PANELS SUBJECTED TO AXIAL COMPRESSION a~ee:E~pers ma~~ egn ggg *~
=MEN IS ~/.'S~i I well ME 2
103 e all 6 10111011111 No mo m Ke 2 102 e 6 I Eae :~~~e on aE OMNI 2 11 10 e NOW 6 4 now 2 iiii~ D iii ~i ~~ : 111100111 1.0 , mill --mll I INI-1111 milim mill 2 4 6/ 2 4 6 e 2 4 6/ 2 4 * / ] 4 * / 1.0 10 10? I0' 104 I co Z WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Buckling-Stress Coefficient, KC FIGURE 3.8B-8
BUCKLING-STRESS COEFFICIENT, KS, FOR UNPRESSURIZED CURVED PANELS SUBJECTED TO SHEAR 104 memo 8 4 ~I/I~~~~MB~~~iill i~~it~~u~uu~utu 2 iiii~~u~~mt~~ttua 103 O 6 ff2f 1_~1~ t ter a ~~~~It~~~rdd Ks 2. Z-2 VI -02 NIIIUU~%~ 102 a 6 4
~~~~ttm~~utuu~ .nr~~tt~utn~,
2 AMEN- t~~iI~W~ .el1~.81U~~1l1~ 10 s~ w~now, mu B 6 _..
~~~ttt~uE~~~uuu~~
~~it1l~l~~ittllt~titulq~~it~
~~tn~a~~t~u~tt~tu~i~~~~tt~
1.4 2 4 6 t to 2 IC= 2 4 61 103 2 4 4 i 104 z WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Buckling-Stress Coefficient, Ks FIGURE 3.8B-9
BUCKLING-STRESS COEFFICIENT, KS, FOR UNPRESSURIZED CURVED PANELS SUBJECTED TO SHEAR 12 11 -pg n r
~~ww w~ri~~~rri nnnn~n nnnnl ~~ .Inn
=nnnn Eonin~/'~nn
~ ~Iln~
r~1~rsaY w~Vpt ,i - -
~~t~~l~ to\LVt11t~~~~ ~Mmmmn MUMM"I ~O>~
nnnn~/~n \'llH nnnnn~nit 'NS0e/ nn~~in~nmm~ tl I~a~ml~~ wl menu
~nnnnli~nnnnlntnnninnn
~nnnlgl~nnnnlrnnnin~
~nlllnl~~1111n{rnllll~
1.4 10 102 103 la z WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Buckling-Stress Coefficient, Ks FIGURE 3.8B-10
WBN APPENDIX 3.8C DOCUMENTATION OF CB&I COMPUTER PROGRAMS 3.8C.1 INTRODUCTION This appendix presents abstracts of the computer programs employed in the design and analysis of the Watts Bar containment vessels. These abstracts explain the purpose of the program and give a brief description of the methods of analysis performed by the program. Analytical derivations are not contained herein, but are in the CB&I Stress Report. 3.8C.2 PROGRAM 1017-MODAL ANALYSIS OF STRUCTURES USING THE EIGEN VALUE TECHNIQUE The purpose of this program is three-fold:
- 1. To calculate the mass and stiffness matrices associated with the structural model.
- 2. To determine the undamped natural periods of the model.
- 3. To calculate the maximum modal responses of the structure; i.e., deflections, shears, and moments.
The stiffness and mass matrices may be required in order to perform a dynamic analysis of the structure. The maximum modal responses may be used to perform a spectral analysis. The program has the following options:
- 1. Vertical translation.
- 2. Torsional modes.
- 3. Soil-structure interaction.
- 4. Liquid sloshing.
- 5. Direct introduction of stiffness and mass matrices.
3.8C.3 PROGRAM 1044-SEISMIC ANALYSIS of VESSEL APPENDAGES Appendages to a vessel may not significantly contribute structurally to the dynamic responses of a model of a vessel. However, appendages can effect the vessel locally by vibrating differently from the model of the vessel at the point of attachment. The response spectrum method of analysis is not a strictly adequate way of obtaining the maximum appendage accelerations since it does not include the possible consequences of near resonance between the vessel model and the appendage model. 3.8C-1
WBN This paper describes the method used to evaluate the maximum elastic differential accelerations between an independently vibrating appendage model and an elastic beam vessel model at the appendage elevation due to known excitations of the elastic beam model. The method involves two distinct steps. Firstly, the necessary time-absolute acceleration records are computed at appendage elevations due to model excitations. Secondly, the maximum differential accelerations between each appendage model and the vessel model at the appendage elevation are obtained. The time-absolute acceleration records at the appendage elevation are computed by use of a step-by-step matrix analysis procedure. The equations of motion for the vessel model are of the form: [M] {} + (AT/) [K] {} + [K] {u} = -[M] {g} where [M] = Mass matrix, order n x n obtained from a modal analysis. [K] = Stiffness matrix, order n x n, obtained from a modal analysis. A = Portion of first mode critical damping for the model T = First mode of the model [M] = A diagonal matrix, order n x n, with diagonal elements corresponding to elements of the mass excited by translational accelerations. {} = n x 1 matrix of relative accelerations between the model base and the n degrees of freedom. {} = n x 1 matrix of velocities corresponding to {} {u} = n x 1 matrix of displacements corresponding to {} {g} = n x 1 matrix of translation base acceleration. n = Degrees of freedom of vessel model. By taking a small time increment (smaller than the smallest period obtained from the model analysis) and letting accelerations vary linearly within the selected increment, the equations of motion can be integrated for the quantities {u}, {}and {} over the expected time increment[1]. The values obtained are superimposed upon the values of these quantities existing at the beginning of the time increment. This process is repeated for the duration of the excitation. The time-absolute acceleration records for each translational degree of freedom are the sums of {} and {g} taken throughout the history of the excitation. 3.8C-2
WBN The second step is similar to the first step. The equation of motion (n = 1) is written for the appendage as a single degree-of-freedom elastic model using the time-absolute acceleration record obtained in Step 1 at the appendage elevation as the excitation. This equation is solved in the same manner used in Step 1. The maximum absolute value of {} obtained is the quantity desired. It is the maximum differential acceleration between the appendage model and the vessel model due to a known excitation of the vessel mode. For any appendage, this two-step procedure should be executed three times. This is required to evaluate normal, tangential and vertical appendage accelerations with respect to a vessel cross-section. 3.8C.4 PROGRAM E1668-SPECTRAL ANALYSIS FOR ACCELERATION RECORDS DIGITIZED AT EQUAL INTERVALS Program E1668 evaluates dynamic response spectra at various periods and presents the results on a printed plot. Given the time-acceleration record, the program numerically integrates the normal convolution time integral for various natural periods and damping ratios. The computed relative displacements, relative and pseudo-relative velocities, and absolute and pseudo-absolute accelerations are tabulated for periods from 0.025 seconds to 1 second. 3.8C.5 PROGRAM 1642-TRANSIENT PRESSURE BEAM ANALYSIS The program was developed to perform the numerical integration required for the transient pressure beam analysis. The pressure transient curve for each compartment is read in and stored as a series of coordinates. At any time instant the total force acting at each compartment is calculated by multiplying the pressures by the corresponding areas of the shell over which they act. Each force is then distributed to the vessel model masses directly above and below. The proportion applied to each mass is based upon their respective distances from the force. For a given increment the program checks the current time, determines the current pressure in each compartment, calculates the current force on each mass and applies a recurrence formula. The deflection values, y(t) and y(t-t) are updated, the current time incremented, and the process is repeated. The values of the orthogonal deflections are stored and also printed out for a prescribed number of times, every ten increments or so, and the equivalent static forces determined. Equivalent static forces are those which produce deflections identical to the calculated kinetic deflections; they are obtained by multiplying the deflection vector by the stiffness matrix. [F] equivalent = [K] [Y] The shears and moments at the particular time are then determined from statics: Q = F M = F
- moment arm 3.8C-3
WBN 2 2 The maximum moments, M = Mx + My Q2x + Q2y and maximum shears, Q = are then printed out at selected locations. In addition the program will also print out an acceleration trace at the mass points. 3.8C.6 PROGRAM E1623-POST PROCESSOR PROGRAM FOR PROGRAM E1374 Program E1623 was written specifically for the TVA Watts Bar Containment Vessels. It performs the following operations:
- 1. Using Fourier data generated by Program El374 (Dynamic Shell Analysis), the summed displacements, forces and stresses found for various points around the shell circumference at each output point on the meridian.
- 2. The maximum of the summed values along with the associated time and azimuth are saved for each elevation and printed out at the end of the problem.
- 3. The following tables are printed:
- 1. Radial deflection, , at each elevation versus azimuth
- 2. Longitudinal force, N, at each elevation versus azimuth
- 3. Longitudinal moment, M, at each elevation versus azimuth
- 4. Circumferential force, N, at each elevation versus azimuth The time basis for these tables is the occurrence of the minimum longitudinal force at the base.
- 4. Ring forces are calculated and then the maximums are pointed out.
- 5. Displacement traces at several elevations can be saved on a tape or disk unit.
- 6. The membrane stress resultants are saved on either a tape or disk unit for input into the buckling check program.
Program E1374 writes the Fourier amplitude results of the fundamental variables (, ' B, , Q, N, M, N) on a labeled tape after each timestop. Program E1623 reads this tape, interpolates to obtain the values at the output times, and calculates the remaining forces and all the stresses. 3.8C-4
WBN The amplitudes are then summed using the following equation: m m f(, , t) = g n (x, t) cos n + h n (x, t) sin n n =1 n =1 where: x = meridinal coordinate t = time gn(x,t) = amplitudes of cosine harmonics hn(x,t) = amplitudes of sine harmonics
= azimuth f(x,,t) = Fourier sum m = maximum number of circumferential waves 3.8C.7 PROGRAM E1374-SHELL DYNAMIC ANALYSIS
- 1. Introduction Program E1374 is CBI's shell dynamic analysis program. Presently, it is capable of extracting eigenvalues and performing undamped transient analyses. Non-axisymmetric loads can be handled through the use of appropriate Fourier series.
The equation of motion for a particular Fourier harmonic n of an undamped system is [Mn] [Un] + [Kn] [Un] = [Pn] where: [Mn] = Mass matrix [Kn] = Stiffness matrix [Pn] = Applied load [Un] = Displacement [Un] = Acceleration Note that all of the above are functions of n. In order to calculate free vibration frequencies and mode shapes the applied load is set equal to zero, [Un] is assumed to be a harmonic function of time, and the eigenvalues and eigenvectors of the resulting equation obtained. If the transient response due to a time-varying load is required, a numerical integration technique is used. 3.8C-5
WBN Since Program E1374 is not set up to handle longitudinal stiffeners, the integration for this portion of the shell is performed using Program 781. The influence values are then converted to stiffness matrix form and stored on disc. After Program E1374 has set up the stiffness matrices for the unstiffened shell, the matrices for segments with stiffening are replaced with the Program 781 matrices from disc. The solution in Program E1374 then continues in the standard manner. This consists of assembling the overall stiffness matrix [Kn] and load vector [Cn], reducing to upper triangular form, and back-substituting. 3.8C.8 PROGRAM E1622-LOAD GENERATION PREPROCESSOR FOR PROGRAM E1374 In order to perform non-axisymmetric analyses on shells, the load must often be defined using Fourier series representation. The purpose of Program E1622 is to calculate and store on magnetic tape a time history of the Fourier pressure amplitudes. The format of this tape is designed specifically for use with Program E1374. In order to calculate the amplitudes of the harmonics several assumptions are made in the program.
- 1. A linear function in the circumferential direction is assumed between given points.
- 2. Only distributed loads are considered.
- 3. The model consists of a cylindrical shell and optional hemispherical top head.
- 4. The pressure has a block type distribution in the longitudinal direction.
- 5. Any initial pressure acting on the shell can be subtracted from the input pressure histories.
- 6. Amplitudes for both sine and cosine terms can be calculated with the user supplying the range of harmonics to be output.
3.8C.9 PROGRAM E1624 SPCGEN-SPECTRAL CURVE GENERATION Program E1624 reads the Fourier amplitudes of the deflection transients stored on magnetic tape from the output of Program E1374. The program calculates the accelerations at uniform time intervals and evaluates the response spectra. From the deflection transient for each harmonic, the acceleration traces are computer generated using three point central difference for the first and last three time steps, and a seven point central difference elsewhere. 3.8C-6
WBN 3.8C.10 PROGRAM 781, METHOD OF MODELING VERTICAL STIFFENERS N = No. of vertical stiffeners around E = Modulus of elasticity The shell shown in Figure 3.8C-l is modeled using 2 layers. The inside layer represents the shell and, therefore, has the normal isotropic material properties. The outer layer, on the other hand is described as an orthotropic material having the following properties. t2 = d bN E 2 = E 2 R E2 = 0 G2 = 0 where: t2 = Thickness of outer layer E2 = Modulus of elasticity of outer layer in longitudinal direction E2 = Modulus of elasticity of outer layer in circumferential direction. G2 = Shear modulus of outer layer. 3.8C.11 PROGRAM 119-CHECK of FLANGE DESIGN This program is used for the design of bolted flanges. The program checks the flange design based on Appendix II of ASME Code, Section VIII. Bolt and flange stresses are computed for both the bolt-up and design conditions. If the bolt and gasket are not overstressed, the computer automatically calculates the required flange thickness or checks any supplied thickness. The minimum gasket width required to prevent crushing, and the maximum pressure that the flange is capable of resisting under the design conditions are automatically calculated. 3.8C.12 PROGRAM 772-NOZZLE REINFORCEMENT CHECK This is a program for checking nozzle reinforcing. It is designed essentially for containment vessels, and adheres to area replacement criteria specified by ASME Section III and VIII. The program does no design work, merely checking the adequacy of pre-selected reinforcing plate dimensions and weld sizes. 3.8C-7
WBN 3.8C.13 PROGRAM 1027-WRC 107 STRESS INTENSITIES AT LOADED ATTACHMENTS FOR SPHERES OR CYLINDERS WITH ROUND OR SQUARE ATTACHMENT This program determines the stress intensities in a sphere or cylinder at a maximum of 12 points around an externally loaded round or square attachment. Stresses resulting from external loads are superimposed on an initial pressure stress situation. The program computes stresses at three levels of plate thicknesses: outside, inside, and centerline of plate. The 12 points investigated are shown in Figure 3.8C-2. Four points at the edge of attachment, at 1/2RT from the edge of attachment and at the edge of reinforcement. The program determines 3 components for each stress intensity:
- 1. X = A normal stress parallel to the vessel's longitudinal axis
- 2. = A normal stress in a circumferential direction
- 3. = A shear stress The program has an option, whereby the influence coefficients can be calculated directly. The program uses the methodology from the "Welding Research Council Bulletin #107", of December 1968. Additionally, the program contains extrapolations of the curves for cylinders in WRC 107 for gamma up to 600. It should be noted that the use of the program requires complete familiarity with WRC 107 publication.
3.8C.14 PROGRAM 1036M-STRESS INTENSITIES IN JUMBO INSERT PLATES This program determines the stress intensities in a "Jumbo" insert plate (a reinforcing plate with multiple penetrations) in a cylindrical vessel at 8 points around one of these penetrations due to the loading on that penetration plus the loadings on the 4 adjacent penetrations all as superimposed on an initial stress situation. It does this at three levels of plate thickness: outside, inside, and centerline of plate. The 8 points investigated are shown in Figure 3.8C-3. The 4 points on radius R are at the junction of the penetration and the insert plate. The other 4 points are other points of interest; normally, they will be at the midpoints in the clear space between penetrations or at the edge of reinforcing. Although 5 penetrations are considered, each point is analyzed as though it were only influenced by 2 (the central penetration plus the penetration on the same axis as the point concerned). The program also determines 3 components for each stress intensity: 1 x = a normal stress parallel to the vessel's longitudinal axis 2 = a normal stress in a vessel's circumferential direction 3 = a shear stress 3.8C-8
WBN Each of these is composed of 3 subcomponents:
- 1. One due to the central penetration's loading
- 2. One due to the loading on the next adjacent penetration
- 3. An initial stress component (input)
The program has an option whereby the penetration loads will be considered reversible or nonreversible in direction. Under the reversible option, (see Figure 3.8C-4) only the data associated with the most severe loading situations is printed out. Most of the analysis and notation used in the program is taken directly from the "Welding Research Council Bulletin #107" of December 1968. Use of the program requires complete familiarity with this publication. The analysis in WRC 107 is for a single penetration. This program analyzes the several penetrations individually, using WRC 107 techniques verbatim, and then through superposition obtains the composite results. The adjacent penetrations must be on a cardinal line of the central penetration in order to use WRC 107 methods. This has required a very conservative extension of the WRC 107 analysis. WRC 107 analysis applies only to the points on the penetration to shell juncture. This program makes stress determinations at points removed from the junction by fictitiously extending the radius of any penetration to any point at which a stress determination is desired. This disregards the statement in WRC 107 that "these stresses attenuate very rapidly at points removed from the penetration to shell juncture". Furthermore, in some cases, the moment induced stresses at both the juncture and at points removed from the juncture are increased by 20% per discussion in WRC 107. Figure 3.8C-5 shows the cases for the 20% increase and indicates the thickness used for the calculation of the parameters (per WRC 107) and stresses. The program contains extrapolations of the curves in WRC 107 for T up to 600. The program is limited to the domains and range of Figures 1A through 4C in WRC 107 (0 < < 0.5 and 5 < T <600). REFERENCE
- 1. Wilson & Clough, Dynamic Response by Step-By-Step Matrix Analysis 3.8C-9
d b VERTICAL STIFFENER MODEL FIGURE 3. 8C-1 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Vertical Stiffener Model FIGURE 3.8C-1
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WBN APPENDIX 3.8D COMPUTER PROGRAMS FOR STRUCTURAL ANALYSIS Historical Information [Computer programs used for structural analysis meet the TVA Quality Assurance Program for computer software. The following sections are for historical purposes.] Computer programs used for structural analysis and design have been validated by one of the following criteria or procedures:
- a. The following computer programs are recognized programs in the public domain:
Usage Start Program Date: Year Hardware Source AMG032 1965 IBM R&H AMGO33 1965 IBM R&H AMGO34 1965 IBM R&H ANSYS 1972 CDC CDC ASHSD 1969 IBM UCB BASEPLATE II 1982 CDC CDC GENDHK 3 1969 IBM UCB GENSHL 2 1969 IBM FIRL GENSHL 5 1968 IBM FIRL GTSTRUDL 1979 CDC GT NASTRAN (MSC) 1974 CDC CDC SAP IV 1973 CDC UCB SAP IV 1974 IBM USC SDRC FRAME 1977 CDC SDRC PACKAGE SAGS/DAGS SPSTRESS 1977 CDC CDC STARDYNE 1977 CDC CDC STRESS 1970 EG CDC STRUDL (V2 M2) 1972 IBM ICES STRUDL (Rel. 2.6) 1974 IBM MCAUTO (Dynal) STRUDL (Rel. 4.0) 1975 IBM MCAUTO STRUPAK PACKAGE 1971 CDC TRW MAP2DF/SAP2DF SUPERB 1977 CDC CDC WELDDA 1983 CDC CDC WERCO 1978 CDC AAA 3.8D-1
WBN All programs on IBM hardware are run under the MVS operating system, on either a 370/165 machine or a 360150 machine. All programs on CDC hardware are run under the SCOPE 3.3 operating system on a 6600 machine. The following abbreviations are used for program sources: CDC - Control Data Corporation, Minneapolis, MI FIRL - Franklin Institute Research Labs, Philadelphia, PA GT - Georgia Institute of Technology, Atlanta, GA ICES - Integrated Civil Engineering System, Worcester, MA MCAUTO - McDonnell-Douglas Automation Company, St. Louis, MO R&H - Rohm & Haas Company, Huntsville, AL SDRC - Structural Dynamics Research Corporation, Cincinnati, OH TRW - TRW Systems Group, Redondo, CA UCB - University of California, Berkely, CA USC - University of Southern California, Los Angeles, CA AAA - AAA Technology and Specialties Co., Inc., Houston, TX
- b. The following programs have been validated by comparison with a program in the public domain:
RESPONSE FOR EARTHQUAKE AVERAGING SPECTRAL RESPONSE Summary comparisons of results for these computer programs are provided in Figures 3.8D-l and 3.8D-2.
- c. The following programs have been validated by comparison with hand calculations:
BIAXIAL BENDING - USD CONCRETE STRESS ANALYSIS DL42 PLTDL42 THERMCYL TORSIONAL DYNANAL PNA100 The following programs have been validated by comparison with analytical results published in the technical literature: BAP222 DYNANAL ROCKING DYNANAL Summary comparison of results for these computer programs are provided in Tables 3.8D-l through 3.8D-10. 3.8D-2
WBN TABLE 3.8.D-1 BIAXIAL BENDING - USD Moment Capacity (FT-KIPS) MX MY Hand Hand Calculations Program Calculations Program 0 0 409 408 601 603 287 285 850 850 164 165 911 909 77 76 933 932 0 0 Comparison of hand calculations with BIAXIAL BENDING - USD for the moment capacities of a reinforced concrete section for a given direct load.
WBN TABLE 3.8D-2 CONCRETE STRESS ANALYSIS Concrete Compression Stress (psi) Hand Program Calculations 436. 436. Steel Tensile Stress (psi) Hand Calculations Program 1 -3833 -3830 2 -2238 -2234 3 - 644 - 639 4 950 957 5 2417 2419 6 3884 3881 7 5478 5477 8 6275 6275 9 11053 11061 Comparison of hand calculations with CONCRETE STRESS ANALYSIS for reinforced concrete beam with 9 rows of steel, subject to combined load of moment and axial force.
WBN TABLE 3.8D-3 THERMCYL Dead Maximum Concrete Steel Load Compression Stress Tensile Stress (psi) (psi) (psi) Hand Program Hand Program Calculations Calculations 0 770.8 770.9 12,948 12,950 10 848.8 848.3 12,285 12,290 100 1313.0 1316.0 8,336 8,311 1000 2795.0 2793.0 -5,010 -4,990 Comparison of hand calculations with THERMCYL results for stresses in reinforced concrete thin-walled cylinder with non-linear temperature distribution across wall thickness and varying dead load axial stress.
WBN TABLE 3.8.D-4 TORSIONAL DYNANAL Pure Torsion Modal Frequencies Mode Frequency (RAD./SEC.) No. Hand Calculations Program 1 2810 2814 2 8430 8430 Comparison of hand calculations with TORSIONAL DYNANAL results for torsional modes of vibration of a thin-walled steel half-tube.
WBN TABLE 3.8D-5 DYNANAL Modal Periods Including Effects of Flexural and Shear Deformations Mode Period (SEC) No. Published Program Results 1 1.48 1.50 2 .425 .430 3 .216 .222 4 .149 .157 5 .114 .124 Comparison of DYNANAL with analytical procedure presented in Engineering Vibrations, L. S. Jacobsen and R. S. Ayre, McGraw-Hill, 1958, Chapter 10, Modal Analysis of 200 Ft. shear-wall building including effects of flexural and shear deformations.
WBN TABLE 3.8D-6 ROCKING DYNANAL Modal Frequencies of Lumped-Mass Shear Beam Including effects of Base Rocking Mode Frequency (RAD./SEC.) No. Published Program Results 1 5.155 5.339 2 20.52 19.226 Comparison of ROCKING DYNANAL with Analytical Procedure presented in "Earthquake Stresses in Shear Buildings," M. G. Salvadori, ASCE Transactions, 1953, Paper No. 2666. Modal analysis of lumped-mass shear beam including effects of base rocking.
WBN TABLE 3.8D-7 BAP222 Comparison of BAP222 with analytical procedure presented in A Simple Analysis for Eccentrically Loaded Concrete Sections, L. G. Parker and J. J. Scanion, Civil Engineering, October 1940 Published Results Program Pressure bulb geometry, Z4 12 (in.) 12 (in.) Pressure bulb geometry, Z5 6.41 (in.) 6.41 (in.) Pressure bulb geometry, Z7 3.67 (in.) 3.36 (in.) Concrete pressure force -14.08 (k) -14.48 (k) Anchor load 1 -1.715 (k) -1.65 (k) Anchor load 2 5.34 (k) 5.4 (k) Anchor load 3 -3.22 (k) -3.44 (k) Anchor load 4 3.665 (k) 3.61 (k)
WBN TABLE 3.8D-8 DL42 Comparison of hand calculations with DL42 for the design of a baseplate resisting a given load. Hand Calculations Program Safety factor (0.5 SSE) 3.232 3.234 Safety factor (SSE) 3.878 3.881 Maximum plate moment 10.535 (k-in) 10.526 Effective section modulus 1.261 (in.) 1.261 Minimum plate thickness 0.417 (in.) 0.417
WBN TABLE 3.8D-9 (Sheet 1 of 1) PLTDL42 Comparison of hand calculations with PLTDL42 for the design of a baseplate resisting a given load. Hand Calculations Program Safety factor (0.5 SSE) 3.232 3.234 Safety factor (SSE) 3.878 3.881 Maximum plate moment 10.535 (k-in) 10.526 Effective section modulus 1.261 (in.) 1.261 Minimum plate thickness 0.417 (in.) 0.417 Plate bending stress 8.355 (k/in) 8.348
WBN TABLE 3.8D-10 PNA 100 NOZZLE STRESSES (PEN X-57) NEXT TO SHELL Calc. Case Mode A B C D 1 Program 11,039 16,588 11,224 16,495 Hand 11,036 16,584 11,221 16,491 4 Program 13,074 19,192 12,417 17,974 Hand 13,070 19,187 12,412 17,968 AWAY FROM SHELL Calc. Case Mode A B C D 1 Program 10,358 10,095 10,571 10,330 Hand 10,354 10,090 10,567 10,327 4 Program 12,944 12,621 12,196 11,915 Hand 12,939 12,616 12,190 11,908
RESPONSE FOR EARTHQUAKE AVERAGmG 0 0 0 N I
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- l 0- 1 Comparison of STRUDL with RESPONSE FOR EARTHQUAKE WATTS BAR NUCLEAR PLANT AVmAGmG Comparison of STRUDL with RESPONSE FOR EARTHQUAKE FINAL SAFETY for a normal-mode time-histo!'Y anal.ysis of a lumped-mass AVERAGING for a normal-mode time-history analysis of a lumped-structural model of a nuclear power plant structure ANALYSIS subjected REPORT mass structural model of a nuclear power plant structure subjected to to the 1940 El Centro earthquake N-S ground motion.
the 1940 El Centro earthquake N-S ground motion. Response for Earthquake FIGURE Averaging 3.8D-1 FIGURE 3.8D-1
SPECTRAL RESFONSE 0 0 0 N 0 0 0
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°o .oo 0.20 0-40 o.so o.eo 1 .oo ACCELERATION - G Comparison of STRUDL with SPECTRAL RESPONSE for a response spectrum analysis of a lumped-mass structure.1 modelBAR WATTS ofNUCLEAR a PLANT Comparison nuclearof power STRUDLplant with SPECTRAL structure subjected to the 194-0 El Centro FINAL SAFETY RESPONSE for a response earthquake N-S spectrum ground analysis motion. of a ANALYSIS REPORT lumped-mass structural model of a nuclear power plant structure subjected to the 1940 El Centro earthquake N-S ground motion.
FIGURE 3.BD-2 Spectral Response FIGURE 3.8D-2
WBN APPENDIX 3.8E CODES, LOAD DEFINITIONS AND LOAD COMBINATIONS FOR THE MODIFICATION AND EVALUATION OF EXISTING STRUCTURES AND FOR THE DESIGN OF NEW FEATURES ADDED TO EXISTING STRUCTURES AND THE DESIGN OF STRUCTURES INITIATED AFTER JULY 1979 3.8E.1 Application Codes and Standards
- a. American Concrete Institute (ACI) 318-77, "Building Code Requirements for Reinforced Concrete"
- b. American Institute of Steel Construction (AISC), "Specification for the Design Fabrication, and Erection of Structural Steel for Buildings," 7th edition adopted February 12, 1969, as amended through June 12, 1974 or later editions, except welded construction is in accordance with Item d below.
- c. American Society for Testing and Materials (ASTM) Standards
- d. American Welding Society (AWS), Structural Welding Code, AWS D1.1-72, with Revisions 1-73 and 2-74 except later editions may be used for prequalified joint details, base materials, and qualification of welding procedures and welders.
Visual inspection of structural welds will meet the minimum requirements of Nuclear Construction Issues Group documents NCIG-01 and NCIG-02 as specified on the design drawings or other engineering design output (See Section 3.8.4.1.1, Item 18).
- e. National Fire Protection Association Standard NFPA 13
- f. National Fire Protection Association Standard NFPA 14
- g. National Fire Protection Association Standard NFPA 15
- h. National Fire Protection Association Standard NFPA 24
- i. National Fire Protection Association Standard NFPA 30
- j. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Sections III, VIII, and IX
- k. American Nuclear Standard Institute (ANSI) B31.1, "Power Piping"
- l. AWS D1.1-81, "Structural Welding Code"
- m. AISC-ANSI-N690-1984 "Nuclear Facilities Steel Safety-Related Structures for Design, Fabrication and Erection" 3.8E-1
WBN 3.8E.2 Load Definitions The following terms are used in the load combination equations for structures. Normal loads, which are those loads to be encountered during normal plant operation and shutdown, include: D Dead loads or their related internal moments and forces including any permanent equipment loads; all hydrostatic loads; and earth loads applied to horizontal surfaces. L Live loads or their related internal moments and forces including any movable equipment loads and other loads which vary with intensity and occurrence, such as lateral soil pressure. To Thermal effects and loads during normal operating or shutdown conditions, based on the most critical transient or steady-state condition. Ro Pipe reactions during normal operating or shutdown conditions, based on the most critical transient or steady-state condition. Severe environmental loads include: E Loads generated by the operating basis earthquake (OBE). The term "operating basis earthquake" has the same meaning as "one-half safe shutdown earthquake." W Load generated by the design wind specified for the plant. Extreme environmental loads include: E' Load generated by the safe shutdown earthquake (SSE). The term "safe shutdown earthquake" has the same meaning as the term "design basis earthquake" (DBE). Wt Loads generated by the design tornado specified for the plant. Tornado loads include loads due to the tornado wind pressure, the tornado-created differential pressure, and to tornado-generated missiles. Abnormal loads, which are those loads generated by a postulated high-energy pipe break accident, include: Pa Pressure equivalent static load within or across a compartment generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Ta Thermal loads under thermal conditions generated by the postulated break and including To. 3.8E-2
WBN Ra Pipe reactions under thermal conditions generated by the postulated break and including Ro Yr Equivalent static load on the structure generated by the reaction on the broken high-energy pipe during the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Yj Jet impingement equivalent static load on a structure generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Ym Missile impact equivalent static load on a structure generated by or during the postulated break, as from pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load. In determining an appropriate equivalent static load for Yr, Yj, and Ym, elasto-plastic behavior may be assumed with appropriate ductility ratios, provided excessive deflections will not result in loss of function of any safety-related system. Other loads: C Construction live loads F' Hydrostatic load from the probable maximum flood Fa Flood load generated by a postulated pipe break Concrete capacity: U Concrete section strength required to resist design loads based on the strength design methods described in ACI 318-77. 3.8E.3 Load Combinations - Concrete
- a. For service load conditions, the strength design method is used, and the following load combinations are considered.
(1) U = 1.4 D + 1.7 L (2) U = 1.4 D + 1.7 L + 1.9 E (3) U = 1.4 D + 1.7 L + 1.7 W If thermal stresses due to To and Ro are present, the following combinations are also considered. (1a) U = (0.75) (1.4 D + 1.7 L + 1.7 To + 1.7 Ro) 3.8E-3
WBN (2a) U = (0.75) (1.4 D + 1.7 L + 1.9 E + 1.7 To + 1.7 Ro) (3a) U = (0.75) (1.4 D + 1.7 L + 1.7 W + 1.7 To + 1.7 Ro) Both cases of L having its full value or being completely absent are checked. In addition, the following combinations are considered: (2a') U = 1.2 D + 1.9 E (3a') U = 1.2 D + 1.7 W
- b. For factored load conditions, which represent extreme environmental, abnormal, abnormal/severe environmental and abnormal/extreme environmental conditions, the strength design method is used; and the following load combinations are considered:
(4) U = D + L + To + Ro + E' (5) U = D + L + To + Ro + W t (6) U = D + L + Ta + Ra + 1.5 Pa (7) U = D + L + Ta + Ra + 1.25 Pa + 1.0 (Yr + Yj + Ym) + 1.25 E (8) U = D + L + Ta + Ra + 1.0 Pa + 1.0 (Yr + Yj + Ym) + 1.0 E' For the Additional Diesel Generator Building and structures initiated after July 1979, the three individual tornado-generated loads are combined as follows: W t = Ww W t = Wp W t = Wm Wt = Ww + 0.5Wp W t = Ww + Wm Wt = Ww + 0.5 Wp + Wm where: Wt is the total tornado load, Ww is the tornado wind load, Wp is the tornado-generated pressure differential load, and 3.8E-4
WBN Wm is the tornado missile load. In combinations (6), (7), and (8), the maximum values of Pa, Ta, Ra, Yj, Yr, and Ym, including an appropriate dynamic load factor, is used unless a time-history analysis is performed to justify otherwise. Combinations (5), (7), and (8), are satisfied first without the tornado missile load in (5) and without Yr, Yj, and Ym in (7) and (8). When considering these concentrated loads, local section strength capacities may be exceeded provided there is no loss of function of any safety-related system.
- c. Other load conditions:
(9) U = 1.4 D + 1.4 C (10) U = D + L + F' (11) U = D + Fa 3.8E.4 Load Combinations - Structural Steel
- a. For service load conditions, the elastic working stress design methods of Part 1 of the AISC specifications is used and the following load combinations are considered.
Allowable Stress Load Combinations (1) AISC Allowable* D+L (2) AISC Allowable* D+L+E (3) AISC Allowable* D+L+W
*See Table 3.8E-1 for limiting values If thermal stresses due to To and Ro are present, the following combinations are also considered:
Allowable Stress Load Combinations (1a) 1.5 x AISC Allowable* D + L + To + Ro (2a) 1.5 x AISC Allowable* D + L + To + Ro + E (3a) 1.5 x AISC Allowable* D + L + To + Ro + W
- The allowable stress shall be limited to the values given in Table 3.8E-1.
3.8E-5
WBN Both cases of L having its full value or being completely absent, are checked.
- b. For factored load conditions, the following load combinations are considered.
Allowable Stress Load Combinations (4) 1.6 x AISC Allowable* D + L + To + Ro + E (5) 1.6 X AISC Allowable* D + L + To + Ro + Wt (6) 1.6 x AISC Allowable* D + L + Ta + Ra + Pa (7) 1.6 x AISC Allowable* D + L + Ta + Ra + Pa
+ 1.0 (Yj + Yr + Ym) + E Allowable Stress Load Combinations (8) 1.7 x AISC Allowable* D + L + Ta + Ra + Pa
+ 1.0 (Yj + Yr +Ym) + E' (9) 1.6 x AISC Allowable* D + Fa (10) 1.6 x AISC Allowable* D + L + E (11) 1.6 x AISC Allowable* D + L + Wt
- If thermal loads are not present, the allowable stress shall be limited to the values given in Table 3.8E-1.
Evaluations of miscellaneous and structural steel designed prior to July 1979, may be performed using load combinations (2), (10), and (11) unless other specific loads of a significant nature exist, in which case, the appropriate load combinations of Section 3.8E.4 must be considered. The design of modifications must meet the load combinations in Section 3.8E.4. Thermal analyses using linear elastic methods are performed for restrained Category I structures located in high temperature environments. In combinations (6), (7), and (8), the maximum values of Pa, Ra, Ta, Yj, Yr, and Ym, including an appropriate dynamic load factor, was used unless a time-history analysis was performed to justify otherwise. Combinations (5), (7), and (8) were first satisfied without the tornado missile load in (5) and without Yr, Yj, and Ym in (7) and (8). 3.8E-6
WBN TABLE 3.8E-1 (Sheet 1 of 2) LIMITING VALUES OF ALLOWABLE STRESS Shear on Compression Loading Tension on Gross on Combinations Net Section Section Section Bending (1), (2), (3) 0.60Fy 0.40Fy See Note 1 See Note 2 (1a), (2a), (3a) 0.90Fy 0.90 Fy See Note 3 0.90Fy (4) through (9) 3 Note 1 - Varies with slenderness ratio, see AISC "Manual of Steel Construction," 7th Edition, Table 1-36, Page 5-84. Note 2 - Varies, see Section 1.5.1.4, "Bending", of Item 3.8E.1.b Note 3 - Varies with slenderness ratio. The allowable stress was obtained from AISC Specification Section 1.5, using formula 1.5-1 or 1.5-2 and 1.5-3 with modifications, as shown below: Main and secondary K 2 ( r ) members where K / r C c : Fa = 0.9Fy 1 (Formula A) 2C c2 0.9 2 E Main members where C c < K / r < 200 : Fa = 2 (Formula B) K r Fa [byFormula(A)or (B)] Secondary members where 120 < K / r 200 : Fas = 1.6 200r Where: 2 2 E Cc = F Fy E = Modulus of elasticity of steel
WBN TABLE 3.8E-1 (Sheet 2 of 2) LIMITING VALUES OF ALLOWABLE STRESS Fa = Axial compressive stress permitted in the absence of bending moment (kips per square inch) Fas = Axial compressive stress, permitted in the absence of bending moment, for bracing and other secondary members (kips per square inch) Fy = Specified minimum yield stress of materials (kips per square inch) K = Effective length factor
= Actual unbraced length (inches) r = Governing radius of gyration (inches)
WBN 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9.1 General Topic for Analysis of Seismic Category I ASME Code and Non-Code Items 3.9.1.1 Design Transients Transients used in the design and fatigue analysis for Westinghouse supplied ASME Code Class 1 components and RCS components are discussed and presented in Section 5.2.1.5. Specifically, the transients are identified for Class 1 components in Tables 5.2-2 and 5.2-3. The transients used in the design and analysis of RCS components are identified in Table 5.2-2. 3.9.1.2 Computer Programs Used in Analysis and Design 3.9.1.2.1 Other Than NSSS Systems, Components, Equipment, and Supports
- 1. The following computer programs are used in piping analyses:
- a. TPIPE Program - TPIPE is a special purpose computer program capable of performing static and dynamic linear elastic analyses of power-related piping systems. The dynamic analysis option includes: (1) frequency extraction, (2) response spectrum, (3) time history modal superposition, and (4) time history direct integration methods.
In addition to these basic analysis capabilities, the program can perform an ASME Section III, Class 1, 2, or 3 stress evaluation and perform thermal transient heat analysis to provide the linear thermal gradient, T1, nonlinear thermal gradient, T2, and gross discontinuity expansion difference, a Ta - b Tb, required for a Class 1 stress evaluation. This program is owned and maintained by TVA. It has been fully verified and documented and was compared with PISOL, SAP IV, PIPSD, STARDYNE, and SUPERPIPE with excellent correlation. These programs are well recognized and utilized throughout the industry. It is maintained and updates are verified in accordance with the TVA Quality Assurance Program for Computer Software.
- b. Post Processors - The post processors are used in performing the stress evaluations and support load calculations made in the analysis of piping systems.
The programs use moment, force, and deflection data generated by TPIPE. A stress evaluation is made for each joint on the analysis model. The appropriate stress intensifications/stress indices according to the ASME Section III code are utilized in evaluating stresses for the Normal, Upset, Emergency, and Faulted Conditions. Pipe rupture limits and active valve limits are also evaluated. The allowed stress difference for pipe lug attachments and the lug load is calculated for each load condition. 3.9-1
WBN Support and anchor design loads are calculated for each support to meet the requirements given in Section 3.9.3.4.2.
- c. The following computer programs are also used for piping analysis:
Program Source Program Description ME-101 BECHTEL Linear elastic analysis of piping systems - Bechtel Western Power Corp San Francisco, CA. ANSYS SWANSON General purpose finite element program - Swanson Analysis Systems, Inc. Houston, PA.
- 2. The following computer programs are used in support design and equipment/component analysis.
ACRONYM PROGRAM DESCRIPTION FAPPS (ME150) Frame Analysis For Pipe Supports SMAPPS (ME152) Standard Frame Analysis For Small Bore Pipe Supports MAPPS (ME153) Miscellaneous Applications For Pipe Supports IAP Integral Welded Attachments CONAN Allowable Tensile Load For Anchor Bolt Group With Shear Cone Overlap BASEPLATE II Finite Element Analysis Of Base Plates And Anchor Bolts GT STRUDL Structural Analysis Program CASD TVA Computer Aided Support Design Program SUPERSAP Structural Finite Element Analysis Program ANSYS Structural Finite Element Analysis Program STARDYNE Structural Analysis Program 3.9.1.2.2 Programs Used for Category I Components of NSSS Computer programs that Westinghouse uses in analysis to determine structural and functional integrity of Seismic Category I systems, components, equipment and supports are presented in WCAP-8252, Revision 1[1] and WCAP-8929 [10]. 3.9-2
WBN 3.9.1.3 Experimental Stress Analysis No experimental stress analysis was used per se, for the reactor internals. However, Westinghouse makes extensive use of measured results from prototype plants and various scale model tests as discussed in the following Sections 3.9.2.4, 3.9.2.5, and 3.9.2.6. 3.9.1.4 Consideration for the Evaluation of the Faulted Condition 3.9.1.4.1 Subsystems and Components Analyzed by Westinghouse The analytical methods used to evaluate stresses for ASME Class 1 systems and components are presented in Section 5.2.1.10. The results of the analyses are documented in the stress reports that describe the system or component. For reactor internals the faulted condition was evaluated based on a non-linear elastic system analysis and conforms to the requirements of Appendix F of the ASME Code Section III. Analytical methods are described in Section 3.9.2.5. 3.9.1.4.2 Subsystems and Components Analyzed by TVA
- 1. Piping Systems - The methods employed in the analysis of ASME Class 1 and Class 2/3 piping systems are elastic analytical methods as described by the equations of Sections NB-3600 and NC-3600 of the ASME Code.
The faulted condition stress limits specified for Class 1 and Class 2/3 systems are in compliance with the elastic method limits set forth in Appendix F subsection F-1360 of the ASME Section III Code.
- 2. Piping System Supports - The methods employed in the analysis of ASME Code Classes 1, 2, and 3 piping system supports are as follows:
- a. Linear Type - Elastic methods as described by Part I of the AISC, "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings,"
February 12, 1969. (Supplements 1, 2, & 3) (Later edition of the AISC code may be utilized when design safety is not compromised.)
- b. Standard Components - Elastic or load-rated methods as described by Manufacturers' Standardization Society (MSS) SP-58, 1967 edition, "Pipe Hangers and Supports."
3.9-3
WBN The faulted condition stress limits for Class 1, 2, and 3 pipe supports are specified in Section 3.9.3.4.2. For linear supports these faulted condition limits meet the intent and requirements of the elastic method limits set forth in Appendix F, subsection F-1320 or F-1370, of the ASME Section III Code. See Section 3.9.3.4.2. For standard components, the allowable stresses or load ratings of MSS-SP-58 are based on a factor of safety of five based on normal operating conditions. Upset, emergency, faulted, and test conditions were evaluated using Table 3.9-21. This low allowable stress is adequate to assure that active components are properly supported for faulted conditions.
- 3. Mechanical Equipment No plastic instability allowable limits given in ASME Section III have been used when dynamic analysis is performed. The limit analysis methods have the limits established by ASME Section III for Normal, Upset and Emergency Conditions. For these cases, the limits are sufficiently low to assure that the elastic system analysis is not invalidated. For ASME Code Class 1 mechanical equipment, the stress limits for faulted loading conditions are specified in Sections 3.9.3.1.2 and 5.2. For ASME Code Class 2 and 3 mechanical equipment the stress limits for faulted loading conditions are specified in Section 3.9.3.1.2. These faulted condition limits are established in such a manner that there is equivalence with the adopted elastic system analysis. Particular cases of concern are checked by readjusting the elastic system analysis.
- 4. Mechanical Equipment Supports The stress limits for the faulted loading condition of mechanical equipment supports are given in Section 3.9.3.4.1 of Westinghouse's scope of supply, and Section 3.9.3.4.2 for TVA's scope of supply.
3.9.2 Dynamic Testing and Analysis 3.9.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping ASME Code Section III, Subparagraph NB-3622.3, "Vibration," requires that vibration effects in piping systems shall be visually observed and where questionable shall be measured and corrected as necessary. The preoperational piping dynamic effects test program at this plant is as follows:
- a. The dynamic (steady state and transient) behavior of safety related piping systems designated as ASME Class 1, 2, and 3 is observed during the preoperational testing program. Sample and instrument lines beyond the root valves are normally not included.
Also included in the program are those portions of ANSI B31.1 piping which has a potential to exhibit excessive vibrations. 3.9-4
WBN
- b. Preoperational tests involving critical piping systems will be in compliance with Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors."
- c. For the piping systems discussed in Item a., visual observation of the piping will be performed by trained personnel during predetermined, steady-state and transient modes of operation. The maximum point(s) of representative vibration, as determined by the visual observation, will be instrumented and measurement will be taken to determine actual magnitudes, if it is judged to be excessive.
- d. The allowable criteria for measurements shall be either a maximum half-amplitude displacement or velocity value based on an endurance limit stress as defined in the ASME B&PV Code (refer to Section 3.7.3.8.1).
- e. Should the measured magnitudes actually exceed the allowable, corrective measures will be performed for the piping system. Any new restraints, as required by corrective measures, will be incorporated into the piping system analysis.
- f. The flow mode which produced the excessive vibrations will be repeated to assure that vibrations have been reduced to an acceptable level.
- g. The flow modes to which the system components will be subjected are defined, in general terms, in the preoperational test program.
- h. Vibration measurements will also be taken on the vital pumps at baseline and on a periodic basis so that excessive vibration can be corrected early in the program and/or detected if it gradually becomes a problem.
- i. Vibrations of the affected portions of the main steam system during MS isolation valve trip will be tested and the results will be evaluated.
- j. Thermal expansion tests will be conducted on the following piping systems:
Reactor Coolant System (RCS) Main Steam Steam Supply to Auxiliary Feedwater Pump Turbine Main Feedwater Pressurizer Relief Line RHR in Shutdown Cooling Mode Steam Generator Blowdown Safety Injection System (those lines adjoining RCS which experience temperature > 200°F) Auxiliary Feedwater CVCS (Charging line from Regen. Hx to RCS, Letdown Line from RCS to Letdown Hx) 3.9-5
WBN During the thermal expansion test, pipe deflections will be measured or observed at various locations based on the location of snubbers and hangers and expected large displacement. One complete thermal cycle (i.e., cold position to hot position to cold position ) will be monitored. For most systems, the thermal expansion will be monitored at cold conditions and at normal operating temperature. Intermediate temperatures are generally not practical due to the short time during which the normal operating temperature is reached. For the RCS and the main steam system, measurements will be made at cold, 250°F, 350°F, 450°F and normal operating temperatures. Acceptance criteria for the thermal expansion test verify that the piping system is free to expand thermally (i.e., piping does not bind or lock at spring hangers and snubbers nor interfere with structures or other piping), and to confirm that piping displacements do not exceed design limits, as described by ASME Section III (i.e., the induced stresses do not exceed the sum of the basic material allowable stress at design temperature and the allowable stress range for expansion stresses). If thermal motion is not as predicted, the support system will be examined to verify correct function or to locate points of binding of restraints. If binding is found, the restraints will be adjusted to eliminate the unacceptable condition or reanalyzed to verify that the existing condition is acceptable. 3.9.2.2 Seismic Qualification Testing of Safety-Related Mechanical Equipment Design of Category I mechanical equipment to withstand seismic, accident, and operational vibratory loadings is provided either by analysis or dynamic testing. Generally tests are run with either of the following two objectives:
- 1. To obtain information on parts or systems necessary to perform the required analysis, or
- 2. To prove the design (stress or operability) adequacy of a given equipment or structure without performing any analysis of this particular equipment or structure.
The need for the first type of tests is dictated by lack of information on some of the inputs vital to the performance of an analysis. These tests can be either static (to obtain spring constants) or dynamic (to obtain impedance characteristics). The need for the second type of test is mainly dictated by the complexity of the structure/equipment under design. This vibration testing is usually performed in a laboratory or shop on a prototype basis, using various sources of energy. 3.9-6
WBN For general seismic qualification requirements for mechanical and electrical equipment, see Section 3.7.3.16. Laboratory vibration testing can be conducted by employing various forms of shakers, the variation depending on the source of the driving force. Generally, the primary source of motion may be electromagnetic, mechanical, or hydraulic-pneumatic. Each is subject to inherent limitations which usually dictate the choice. To properly simulate the seismic disturbance, the waveform must be carefully defined. The waveform seen by a given piece of equipment depends on:
- 1. The earthquake motion specified for a given site.
- 2. The soil-structure interaction.
- 3. The building in which the component is housed.
- 4. The floor on which the equipment is located.
- 5. The support and attachments to the equipment.
Components located on rocks or on stiff lower floors of buildings founded on rock are subjected to random-type vibrations. Components located on the upper floors of flexible buildings, in flexible subsystems, or in buildings on soft foundations are roughly subjected to sine beats with a frequency close to fundamental frequency of the building or subsystem. In cases where random vibration inputs are used, extreme care is paid to the selection of random forcing functions having frequency content and energy conservatively approaching those of the ground or buildings motion caused by the specified earthquake(s). The most common and readily available vibration testing facilities could only carry simple harmonic motion. By analytical comparison with time history response obtained with a number of real earthquake motions, it has been found that these time histories can be approximately simulated with wave forms having the shape of sine beats with 5 or 10 cycles per beat, a frequency equal to the component natural frequencies, and maximum amplitude equal to the maximum seismic acceleration to which the component needs to be qualified. For equipment located on building floors, the maximum seismic input acceleration is the maximum floor acceleration. This is obtained from the dynamic analysis of the building or from the appropriate floor response spectrum at the zero period of the equipment. 3.9-7
WBN The above procedure adheres closely to the IEEE 344-1971 "IEEE Guide for Seismic Qualification of Class 1 Electric Equipment for Nuclear Power Generating Stations." This standard was specified for equipment for the Watts Bar Nuclear Plant contracted for up to September 1, 1974. New contracts after this date specified IEEE 344-1975 "IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations." The first test to the IEEE 344-1975 was run in March 1974 on 6.9 kV switch gear. On local panels, test qualification to both standards was used because some instruments and controls mounted there on were procured to each version. This one test revealed that the 1971 version of IEEE 344 was the more severe. As an example, seismic qualification and the demonstration of operability of active Class 2 and 3 pumps, active Class 1, 2, or 3 valves, and their respective drives, operators and vital auxiliary equipment is shown by satisfying the criteria given in Section 3.9.3.2. Other active mechanical equipment will be shown operable by either testing, analysis or a combination of testing and analysis. The operability programs implemented on this other active equipment will be similar to the program described in Section 3.9.3.2 for pumps and valves. Testing procedures similar to the procedures outlined in Section 3.10 for electrical equipment will be used to demonstrate operability if the component is mechanically or structurally complex such that its response cannot be adequately predicted analytically. Analysis may be used if the equipment is amen-able to modeling and dynamic analysis. Inactive Seismic Category I equipment will be shown to have structural integrity during all plant conditions in one of the following manners: 1) by analysis satisfying the stress criteria applicable to the particular piece of equipment, or 2) by test showing that the equipment retains its structural integrity under the simulated seismic environment. A list of Category I mechanical equipment and the original method of qualification is provided in the Table 3.7-25. 3.9.2.3 Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady-State Conditions The vibration characteristics and behavior due to flow induced excitation are very complex and not readily ascertained by analytical means alone. Reactor components are excited by the flowing coolant which causes oscillatory pressures on the surfaces. The integration of these pressures over the applied area should provide the forcing functions to be used in the dynamic analysis of the structures. In view of the complexity of the geometries and the random character of the pressure oscillations, a closed form solution of the vibratory problem by integration of the differential equation of motion is not always practical and realistic. Therefore, the determination of the forcing functions as a direct correlation of pressure oscillations can not be practically performed independently of the dynamic characteristics of the structure. The main objective, then, is to establish the characteristics of the forcing functions that essentially determine the response of the structures. By studying the dynamic properties of the structure from previous analytical and experimental work, the characteristics of the forcing function can be deduced. These studies indicate that the most important forcing functions are flow turbulence, and pump-related excitation. The relevance of such excitations depends on many factors such as type and location of component and flow conditions. 3.9-8
WBN The effects of these forcing functions have been studied from test runs on models, prototype plants and in component tests. [2,4,5] The Indian Point Unit 2 plant has been established as the prototype for four-loop plant internals verification program and was fully instrumented and tested during initial startup.[4] In addition, the Sequoyah Unit 1 and Trojan Nuclear Plants have also been instrumented to provide prototype data applicable to Watts Bar.[5] Although the Watts Bar plant is similar to Indian Point Unit 2, significant differences are the modifications resulting from the use of 17 x 17 fuel, the replacement of the annular thermal shield with neutron shielding panels, and reactor vessel barrel/baffle upflow flow design. These differences are addressed below.
- 1. 17 x 17 Fuel The only structural changes in the internals resulting from the design change from the 15 x 15 to the 17 x 17 fuel assembly are the guide tube and control rod drive line. The new 17 x 17 guide tubes are stronger and more rigid, hence they are less susceptible to flow induced vibration. The fuel assembly itself is relatively unchanged in mass and spring rate, and thus no significant deviation is expected from the 15 x 15 fuel assembly vibration characteristics.
- 2. Neutron Shielding Pads Lower Internals The primary cause of core barrel excitation is flow turbulence, which is not affected by the upper internals [3]. The vibration levels due to core barrel excitation for Trojan and Watts Bar both having neutron shielding pads, are expected to be similar. Since Watts Bar has greater velocities than Trojan, vibration levels due to the core barrel excitation is expected to be somewhat greater than that for Trojan (proportional to flow velocity raised to a small power). However, scale model test results and preliminary results from Trojan show that core barrel vibration of plants with neutron shielding pads is significantly less than that of plants with thermal shields. This information and the fact that low core barrel flange stresses with large safety margins were measured at Indian Point Unit 2 (thermal shield configuration) lead to the conclusion that stresses approximately equal to those of Indian Point Unit 2 will result on the Watts Bar internals with the attendant large safety margins.
3.9-9
WBN
- 3. Reactor Vessel Barrel/Baffle Upflow Conversion The upflow conversion consists of changes to the reactor vessel components, which are to plug the core barrel inlet flow holes and to provide holes in the top former plate.
These modifications change the flow path from being down flow to upflow between the core barrel and the baffle plate and increase core bypass flow by 1.5%. Changing the flow path reduces the pressure differential across the baffle plate, eliminating the jetting of coolant between the joints between the baffle plates. Although defined as a difference between Indian Point 2 and Watts Bar internals, the conversion of the Watts Bar internals to the upflow configuration has no direct impact on the reactor core system under earthquake conditions. Therefore, the fuel assembly structural integrity during a seismic event is not affected by the modification. The potential effects due to the LOCA contribution, as a result of the upflow modification, has been demonstrated by evaluation that the impact of the change in forces from the initial down flow design to upflow are insignificant. Therefore, the modifications associated with the upflow conversion do not increase the seismic or LOCA induced loads significantly compared to that of the downflow design, and the fuel assembly structural integrity and coolable geometry are maintained. This issue has been reviewed and approved by the NRC.[11 & 12] 3.9.2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals Because the Watts Bar reactor internals design configuration is well characterized, as was discussed in Section 3.9.2.3, it is not considered necessary to conduct instrumented tests of the Watts Bar plant hardware. The requirements of Regulatory Guide 1.20 will be met by conducting the confirmatory preoperational testing examination for integrity per Paragraph D, of Regulatory Guide 1.20, "Regulation for Reactor Internals Similar to the Prototype Design." This examination will include some 34 points for Unit 1 (Figure 3.9-1a and 29 points for Unit 2 (Figure 3.9-1b) with special emphasis on the following areas.
- 1. All major load-bearing elements of the reactor internals relied upon to retain the core structure in place.
- 2. The lateral, vertical and torsional restraints provided within the vessel.
- 3. Those locking and bolting devices whose failure could adversely affect the structural integrity of the internals.
- 4. Those other locations on the reactor internal components which are similar to those which were examined on the prototype Indian Point Unit 2 design.
3.9-10
WBN
- 5. The inside of the vessel will be inspected before and after the hot functional test, with all the internals removed, to verify that no loose parts or foreign material are in evidence.
A particularly close inspection will be made on the following items or areas using a 5X or 1OX magnifying glass or penetrant testing where applicable.
- 1. Lower Internals
- a. Upper barrel to flange girth weld.
- b. Upper barrel to lower barrel girth weld.
- c. Upper core plate aligning pin. Examine bearing surfaces for any shadow marks, burnishing, buffing or scoring. Inspect welds for integrity.
- d. Irradiation specimen guide screw locking devices and dowel pins. Check for lockweld integrity.
- e. Baffle assembly locking devices. Check for lock weld integrity.
- f. Lower barrel to core support girth weld.
- g. Neutron shield panel screw locking devices and dowel pin cover plate welds.
Examine the interface surfaces for evidence of tightness and for lock weld integrity.
- h. Radial support key welds.
- i. Insert screw locking devices. Examine soundness of lock welds.
- j. Core support columns and instrumentation guide tubes. Check the joints for tightness and soundness of the locking devices.
- k. Secondary core support assembly screw locking devices for lock weld integrity.
- l. Lower radial support keys and inserts. Examine for any shadow marks, burnishing, buffing or scoring. Check the integrity of the lock welds. These members supply the radial and torsional constraint of the internals at the bottom relative to the reactor vessel while permitting axial and radial growth between the two. One would expect to see, on the bearing surfaces of the key and keyway, burnishing, buffing or shadow marks which would indicate pressure loading and relative motion between the two parts. Some scoring of engaging surfaces is also possible and acceptable.
3.9-11
WBN
- m. Gaps at baffle joints. Check for gaps between baffle and top former and at baffle to baffle joints.
- 2. Upper Internals
- a. Thermocouple conduits, clamps and couplings. (Unit 1 Only)
- b. Guide tube, support column, orifice plate, and thermocouple assembly locking devices. (Unit 1 Only)
- c. Support column (Units 1 and 2) and thermocouple conduit assembly clamp (Unit 1 Only) welds.
- d. Upper core plate alignment inserts. Examine for any shadow marks, burnishing, buffing or scoring. Check the locking devices for integrity of lock welds.
- e. Thermocouple conduit gusset and clamp welds (where applicable). (Unit 1 Only)
- f. Thermocouple conduit end-plugs. Check for tightness. (Unit 1 Only)
- g. Guide tube enclosure welds, tube-transition plate welds and card welds.
- h. Guide tube, support column, orifice plate, and flow restrictor locking devices.
(Unit 2 Only) Acceptance standards are the same as required in the shop by the original design drawings and specifications. During the hot functional test, the internals will be subjected to a total operating time at greater than normal full-flow conditions (four pumps operating) of at least 240 hours. This provides a cyclic loading of approximately 107 cycles on the main structural elements of the internals. In addition there will be some operating time with only one, two and three pumps operating. When no signs of abnormal wear, no harmful vibrations are detected and no apparent structural changes take place, the four-loop core support structures are considered to be structurally adequate and sound for operation. 3.9.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions 3.9.2.5.1 Design Criteria The basic requirement of any LOCA including the double-ended severance of a reactor coolant pipe, is that sufficient integrity be maintained to permit the safe and orderly shutdown of the reactor. This implies that the core must remain essentially intact and the deformations of the internals must be sufficiently small so that primary loop flow, and particularly, adequate safety injection flow, is not impeded. The ability to insert control rods, to the extent necessary, to provide shutdown following the accident must be maintained. Maximum allowable deflection limitations are established for those regions of the internals that are critical for plant shutdown. The allowable and no loss of function deflection limits under dead weight loads plus the maximum potential earthquake and/or blowdown excitation loads are presented in Table 3.9-5. These limits have been established by correlating experimental and analytical results. 3.9-12
WBN With the acceptance of Leak-Before Break by NRC, References [6][7][8][9][10][11] and [12] of Section 3.6B.1, the dynamic effects of main coolant loop piping are no longer considered in the design basis analysis. Only the dynamic effects of the next most limiting breaks of auxiliary lines need to be considered; and consequently the components will experience considerably less loads and deformations than those from the main loop line breaks. 3.9.2.5.2 Internals Evaluation The horizontal and vertical forces exerted on reactor internals and the core, following a LOCA, are computed by employing the MULTIFLEX 3.0, which is an enhancement and extension of MULTIFLEX 1.0,[14] NRC reviewed and approved computer code developed for the space-time dependent analysis of nuclear power plants. MULTIFLEX 3.0 has been accepted by NRC for several other applications[15], [16], [17], [18] and also has been extensively used for the LOCA analyses of reactor internals components of numerous other 2, 3, and 4 loop nuclear power plants. 3.9.2.5.3 LOCA Forces Analysis MULTIFLEX[14] is a digital computer program for calculation of pressure, velocity, and force transients in reactor primary coolant systems during the subcooled, transition, and early saturation portion of blowdown caused by a LOCA. During this phase of the accident, large amplitude rarefaction waves are propagated through the system with the velocity of sound causing large differences in local pressures. As local pressures drop below saturation, causing formation of steam, the amplitudes and velocities of these waves drastically decrease. Therefore, the largest forces across the reactor internals due to wave propagation occur during the subcooled portions of the blowdown transient. MULTIFLEX includes mechanical structure models and their interaction with the thermal-hydraulic system. MULTIFLEX Code The thermal-hydraulic portion of MULTIFLEX is based on the one dimensional homogeneous flow model which is expressed as a set of mass, momentum, and energy conservation equations. These equations are quasi-linear first order partial-differential equations that are solved by the method of characteristics. The numerical method employed is the explicit scheme; consequently time steps for stable numerical integration are restricted by sonic propagation. In MULTIFLEX, the structural walls surrounding a hydraulic path may deviate from their neutral positions depending on the force differential on the wall. The wall displacements are represented by those of one-dimensional mass points which are described by the mechanical equations of vibration. MULTIFLEX is a generalized program for analyzing and evaluating thermal-hydraulic-structure system dynamics. The thermal-hydraulic system is modeled with an equivalent pipe network consisting of one-dimensional hydraulic legs which define the actual system geometry. The actual system parameters of length, area, and volume are represented with the pipe network. 3.9-13
WBN MULTIFLEX computes the pressure response of a system during a decompression transient. The asymmetric pressure field in the down-comer annulus region of a PWR can be calculated. This pressure field is integrated over the core support barrel area to obtain total dynamic load on the core support barrel. The pressure distributions computed by MULTIFLEX can also be used to evaluate the reactor core assembly and other primary coolant loop component support integrity. MULTIFLEX evaluates the pressure and velocity transients for locations throughout the system. The pressure and velocity transients are made available to the programs LATFORC and FORCE-2 (described in Reference [14], Appendix A and B), which used detailed geometric descriptions to evaluate hydraulic loading on reactor internals. Horizontal/Lateral Forces - LATFORC LATFORC, described in Reference [14], Appendix A, calculates the lateral hydraulic loads on the reactor vessel wall, core barrel, and thermal shield, resulting from a postulated LOCA in the primary RCS. A variation of the fluid pressure distribution in the down-comer annulus region during the blowdown transient produces significant asymmetrical loading on the reactor vessel internals. The LATFORC computer code is used in conjunction with MULTIFLEX, which provides the transient pressures, mass velocities, and other thermodynamic properties as a function of time. Vertical Forces - FORCE-2 FORCE-2, described in Reference [14], Appendix B, determines the vertical hydraulic loads on the reactor vessel internals. Each reactor component for which force calculations are required is designated as an element and assigned an element number. Forces acting upon each of the elements are calculated summing the effects of:
- 1. The pressure differential across the element.
- 2. Flow stagnation on, and unrecovered orifice losses across, the element.
- 3. Friction losses along the element.
Input to the code, in addition to the MULTIFLEX pressure and velocity transients, includes the effective area of each element on which acts the force due to the pressure differential across the element, a coefficient to account for flow stagnation and unrecovered orifice losses, and the total area of the element along which the shear forces act. 3.9-14
WBN 3.9.2.5.4 Structural Response of Reactor Internals During LOCA and Seismic Conditions Structural Model and Methods of Analysis The response of reactor vessel internals due to an excitation produced by a complete severance of auxiliary loop piping is analyzed. With the acceptance of Leak-Before-Break (LBB) by NRC, Reference [12] of Section 3.6B.1, the dynamic effects of main coolant loop piping no longer have to be considered in the design basis analysis. Only the dynamic effects of the next most limiting breaks of auxiliary lines need to be considered; and consequently the components will experience considerably less loads than those from the main loop line breaks. Assuming that such a pipe break in cold leg occurs in a very short period of time (1 ms), the rapid drop of pressure at the break produces a disturbance that propagates through the reactor vessel nozzle into the down-comer (vessel and barrel annulus) and excites the reactor vessel and the reactor internals. The characteristics of the hydraulic excitation combined with those of the structures affected present a unique dynamic problem. Because of the inherent gaps that exist at various interfaces of the reactor vessel/reactor internals/fuel, the problem becomes that of nonlinear dynamic analysis of the reactor pressure vessel system. Therefore, nonlinear dynamic analyses (LOCA and Seismic) of the reactor pressure vessel system include the development of LOCA and seismic forcing functions. Structural Model Figure 3.9-1 is schematic representation of the reactor pressure vessel system. In this figure, the major components of the system are identified. The reactor pressure vessel system finite element model for the nonlinear time history dynamic analysis consists of three concentric structural sub-models connected by nonlinear impact elements and linear stiffness matrices. The first sub-model, shown in Figure 3.9-2a and 3.9-2b represents the reactor vessel shell and its associated components. The reactor vessel is restrained by four reactor vessel supports (situated beneath alternate nozzles) and by the attached primary coolant piping. Also shown in Figure 3.9-2a is a typical reactor pressure vessel support mechanism. The second sub-model, shown in Figure 3.9-3a represents the reactor core barrel, lower support plate, tie plates, and the secondary support components for Watts Bar Unit 1. These sub-models are physically located inside the first, and are connected by stiffness matrices at the vessel/internals interfaces. Core barrel to reactor vessel shell impact is represented by nonlinear elements at the core barrel flange, upper support plate flange, core barrel outlet nozzles, and the lower radial restraints. The third and innermost sub-model, shown in Figure 3.9-3b represents the upper support plate assembly consisting of guide tubes, upper support columns, upper and lower core plates, and the fuel. The fuel assembly simplified structural model incorporated into the reactor pressure vessel system model preserves the dynamic characteristics of the entire core. For each type of fuel design the corresponding simplified fuel assembly model is incorporated into the system model. The third sub-model is connected to the first and second by stiffness matrices and nonlinear elements. Finally, Figure 3.9-3c shows the reactor pressure vessel system model representation. Analysis Technique The WECAN Computer Code[10] which is used to determine the response of the reactor vessel 3.9-15
WBN and its internals, is a general purpose finite element code. In the finite element approach, the structure is divided into a finite number of discrete members or elements. The inertia and stiffness matrices, as well as the force array, are first calculated for each element in the local coordinates. Employing appropriate transformations, the element global matrices and arrays are assembled into global structural matrices and arrays, and used for dynamic solution of the differential equation of motion for the structure. [M]{U} + [D]{U} + [K]{U} = {F} (1) The WECAN Code solves equation of motions (1) using the nonlinear modal superposition theory. Initial computer runs such as dead weight analysis and the vibration (modal) analyses are made to set the initial vertical interface gaps and to calculate eigenvalues and eigenvectors. The modal analysis information is stored on magnetic tapes, and is used in a subsequent computer runs which solves equation of motions. The first time step performs the static solution of equations to determine steady state solution under normal operating hydraulic forces. After the initial time step, WECAN calculates the dynamic solution of equations of motions and nodal displacements and impact forces are stored on tape for post-processing. The fluid-solid interactions in the LOCA analysis are accounted through the hydraulic forcing functions generated by MULTIFLEX Code. Following a postulated LOCA pipe rupture, forces are imposed on the reactor vessel and its internals. These forces result from the release of the pressurized primary system coolant. The release of pressurized coolant results in traveling depressurization waves in the primary system. These depressurization waves are characterized by a wave front with low pressure on one side and high pressure on the other. The LOCA loads applied to the reactor vessel system for the auxiliary line breaks consist of: (a) reactor internals hydraulic loads (vertical and horizontal, and (b) reactor coolant loop mechanical loads. These loads are calculated individually and combined in a time history manner. Reactor Pressure Vessel Internal Hydraulic Loads Depressurization waves propagate from the postulated break location into the reactor vessel through either a hot leg or a cold leg nozzle. After a postulated cold leg break the depressurization path for waves entering the reactor vessel is through the nozzle which contains the broken pipe and into the region between the core barrel and the reactor vessel (i.e., down-comer region). The initial waves propagate up, around, and down the down-comer annulus, then up through the region circumferentially enclosed by the core barrel, that is, the fuel region. 3.9-16
WBN In the case of cold leg break, the region of the down-comer annulus close to the break depressurizes rapidly but, because of the restricted flow areas and finite wave speed (approximately 3000 feet per second), the opposite side of the core barrel remains at a high pressure. This results in a net horizontal force on the core barrel and the reactor vessel. As the depressurization wave propagates around the down-comer annulus and up through the core, the core barrel differential pressure reduces and, similarly, the resulting hydraulic forces drop. In the case of a postulated break in the hot leg, the wave follows a similar depressurization path, passing through the outlet nozzle and directly into the upper internals region depressurizing the core and entering the down-comer annulus from the bottom exit of the core barrel. Thus, after a reactor pressure vessel outlet nozzle break, the down-comer annulus would be depressurized with very little difference in pressure forces across the outside diameter of the core barrel. A hot leg break produces less horizontal force because the depressurization wave travels directly to the inside of the core barrel (so that the down-comer annulus is not directly involved) and internal differential pressures are not as large as for a cold leg break of the same size. Since the differential pressure is less for a hot leg break, the horizontal force applied to the core barrel is less for hot leg break than for a cold leg break. For breaks in both the hot leg and cold leg, the depressurization waves continue to propagate by reflection and translation through the reactor vessel and loops. The MULTIFLEX computer code, calculates the hydraulic transients within the entire primary coolant system. It considers subcooled, transition, and early two-phase (saturated) blowdown regimes. The MULTIFLEX code employs the method of characteristics to solve the conservation laws, and assumes one-dimensionality of flow and homogeneity of the liquid-vapor mixture. The MULTIFLEX code considers a coupled fluid-structure interaction by accounting for the deflection of constraining boundaries, which are represented by separate spring-mass oscillator system. A beam model of the core support barrel has been developed from the structural properties of the core barrel; in this model, the cylindrical barrel is vertically divided into equally spaced segments and the pressures as well as the wall motions are projected onto the plane parallel to the broken nozzle. Horizontally, the barrel is divided into 10 segments; each segment consists of three separate walls. The spatial pressure variation at each time step is transformed into 10 horizontal forces which act on the 10 mass points of the beam model. Each flexible wall is bounded on either side by a hydraulic flow path. The motion of the flexible wall is determined by solving the global equations of motions for the masses representing the forced vibration of an undamped beam. Reactor Coolant Loop Mechanical Loads The loop mechanical loads result from the release of normal operating forces present in the pipe prior to the separation as well as transient hydraulic forces in the RCS. The magnitudes of the loop release forces are determined by performing a reactor coolant loop analysis for normal operating loads (pressure, thermal, and deadweight). The load existing in the pipe at the postulated break location are calculated and are released at the initiation of the LOCA transient by application of the loads to the broken piping ends. These forces are applied with a ramp time of one millisecond because of the assumed instantaneous break opening time. 3.9-17
WBN In order to obtain the response of reactor pressure vessel system (vessel/internals/fuel), the LOCA horizontal and vertical forces obtained from the LATFORC and FORCE-2 Codes, which were described earlier, together with the loop mechanical loads are applied to the finite element system model shown in Figure 3.9-3c. The transient response of the reactor internals consists of time history nodal displacements and time history impact forces. 3.9.2.5.5 Seismic Analysis The basic mathematical model for seismic analysis is essentially similar to the LOCA model except for some minor differences. In LOCA model, as mentioned earlier, the fluid-structure interactions are accounted though the MULTIFLEX Code; whereas in the seismic model the fluid-structure interactions are included through the hydrodynamic mass matrices in the down-comer region. Another difference between the LOCA and seismic models is the difference in loop stiffness matrices. The seismic model uses the unbroken loop stiffness matrix, whereas the LOCA model uses the broken loop stiffness matrix. Except for these two differences, the reactor pressure vessel system seismic model is identical to that of LOCA model. The horizontal fluid-structure or hydroelastic interaction is significant in the cylindrical fluid flow region between the core barrel and the reactor vessel annulus. Mass matrices with off-diagonal terms (horizontal degrees-of-freedom only) attach between nodes on the core barrel, thermal shield and the reactor vessel. The mass matrices for the hydroelastic interactions of two concentric cylinders are developed using the work of Reference [20]. The diagonal terms of the mass matrix are similar to the lumping of water mass to the vessel shell, thermal shield, and core barrel. The off-diagonal terms reflect the fact that all the water mass does not participate when there is no relative motion of the vessel and core barrel. It should be pointed out that the hydrodynamic mass matrix has no artificial virtual mass effect and is derived in a straight forward, quantitative manner. The matrices are a function of the properties of two cylinders with the fluid in the cylindrical annulus, specifically, inside and outside radius of the annulus, density of the fluid and length of the cylinders. Vertical segmentation of the reactor vessel and the core barrel allows inclusion of radii variations along their heights and approximates the effects of beam mode deformation. These mass matrices were inserted between the selected nodes on the core barrel, thermal shield, and the reactor vessel. The seismic evaluations are performed by including the effects of simultaneous application of time history accelerations in three orthogonal directions. The WECAN computer code, which is described earlier, is also used to obtain the response for the reactor pressure vessel system under seismic excitations. 3.9-18
WBN 3.9.2.5.6 Results and Acceptance Criteria The reactor internals behave as a highly nonlinear system during horizontal and vertical oscillations of the LOCA forces. The nonlinearities are due to the coulomb friction at the sliding surfaces and due to gaps between components causing discontinuities in force transmission. The frequency response is consequently a function not only of the exciting frequencies in the system but also of the amplitude. Different break conditions excite different frequencies in the system. This situation can be seen clearly when the response under LOCA forces is compared with the seismic response. Under seismic excitations, the system response is not as nonlinear as LOCA response because various gaps do not close during the seismic excitations. The results of the nonlinear LOCA and seismic dynamic analyses include the transient displacements and impact loads for various elements of the mathematical model. These displacements and impact loads, and the linear component loads (forces and moments) are then used for detailed component evaluations to assess the structural adequacy of the reactor vessel, reactor internals, and the fuel. The stresses due to the safe shutdown earthquake (SSE) are combined in the most unfavorable manner with the LOCA stresses in order to obtain the largest principal stresses and deflections. These results indicate that the maximum deflections and stress in the critical structures are below the established allowable limits. For transverse excitation of the core barrel, it is shown that the upper core barrel does not buckle during hot leg break. The results also show that the guide tubes will deform well within the limits established experimentally to assure control rod insertion. Seismic deflections of the guide tubes are generally negligible by comparison with the no loss of function limit. 3.9.2.5.7 Structural Adequacy of Reactor Internals Components The reactor internal components of Watts Bar Unit 1 are not ASME Code components. This is due to the fact that Sub-section NG of the ASME Boiler and Pressure Code edition applicable to Watts Bar reactor internals did not include design criteria for the reactor internals since its design preceded Subsection NG of the ASME Code. However, these components were originally designed to meet the intent of the 1971 Edition of Section III of the ASME Boiler and Pressure Vessel Code with addenda through the Winter 1971. As noted previously, that with the acceptance of LBB by NRC, the dynamic effects of the main reactor coolant loop piping are no longer considered in the design basis analysis. Only the dynamic effects of the next most limiting breaks of the auxiliary lines (accumulator line, pressurizer surge line, and RHR lines) are to be considered. It should be noted that LBB discussed in Section 3.6A.2.1.5 also refers to the elimination of pressurizer surge line break from the design basis of Watts Bar Unit 1. Therefore, LOCA response of Watts Bar Unit 1 was determined for the auxiliary line breaks consisting of accumulator line and an RHR line. Consequently, the components experience considerably less loads and deformations than those from the main loop breaks which were considered in the original design of the reactor internals. 3.9-19
WBN Allowable Deflection and Stability Criteria The criterion for acceptability with regard to mechanical integrity analyses is that adequate core cooling and core shutdown must be ensured. This implies that the deformation of reactor internals must be sufficiently small so that the geometry remains substantially intact. Consequently, the limitations established on the reactor internals are concerned principally with the maximum allowable deflections and stability of the components. For faulted conditions, deflections of critical reactor internal components are limited to the values given in Table 3.9-5. In a hypothesized vertical displacement of internals, energy-absorbing devices limit the displacement to 1.25 inches by contacting the vessel bottom head. Core Barrel Response Under Transverse Excitations In general, there are two possible modes of dynamic response of the core barrel during LOCA conditions: a) During a cold leg break the inside pressure of the core barrel is much higher than the outside pressures, thus subjecting the core barrel to outward deflections. b) During a hot leg break the pressure outside the core barrel is greater than the inside pressure thereby subjecting the core barrel to compressive loading. Therefore, this condition requires the dynamic stability check of the core barrel during a hot leg break. (1) To ensure shutdown and cooldown of the core during cold leg blowdown, the basic requirement is a limitation on the outward deflection of the barrel at the locations of the inlet nozzles connected to unbroken lines. A large outward deflection of the upper barrel in front of the inlet nozzles, accompanied with permanent strains, could close the inlet area and restrict the cooling water coming from the accumulators. Consequently, a permanent barrel deflection in front of the unbroken inlet nozzles larger than a certain limit, called no loss of function limit, could impair the efficiency of the ECCS. (2) During the hot leg break, the rarefaction wave enters through the outlet nozzle into the upper internals region and thus depressurizes the core and then enters the down-comer annulus from the bottom exit of the core barrel. This depressurization of the annulus region subjects the core barrel to external pressures and this condition requires a stability check of the core barrel during hot leg break. Therefore, to ensure rod insertion and to avoid disturbing the control rod cluster guide structure, the barrel should not interfere with the guide tubes. Table 3.9-5 summarizes the allowable and no loss of function displacement limits of the core barrel for both the cold leg and hot leg breaks postulated in the main line loop piping. With the acceptance of LBB, the reactor internal components such as core barrel will experience much less loads and deformations than those obtained from main loop piping. 3.9-20
WBN Control Rod Cluster Guide Tubes The deflection limits of the guide tubes which were established from the test data, and for fuel assembly thimbles, cross-section distortion (to avoid interference between the control rood and the guides) are given in Table 3.9-5 Upper Package The local vertical deformation of the upper core plate, where a guide tube is located, shall be below 0.100 inch. This deformation will cause the plate to contact the guide tube since the clearance between the plate and the guide tube is 0.100 inch. This limit will prevent the guide tubes from undergoing compression. For a plate local deformation of 0.150 inch, the guide tube will be compressed and deformed transversely to the upper limit previously established. Consequently, the value of 0.150 inch is adopted as the no loss function local deformation with an allowable limit of 0.100 inch. 3.9.2.6 Correlations of Reactor Internals Vibration Tests With the Analytical Results The dynamic behavior of reactor components has been studied using experimental data obtained from operating reactors along with results of model tests and static and dynamic tests in the fabricators shops and at plant site. Extensive instrumentation programs to measure vibration of reactor internals (including prototype units of various reactors) have been carried out during preoperational flow tests, and reactor operation. From scale model tests, information on stresses, displacements, flow distribution, and fluctuating differential pressures is obtained. Studies have been performed to verify the validity and determine the prediction accuracy of models for determining reactor internals vibration due to flow excitation. Similarity laws were satisfied to assure that the model response can be correlated to the real prototype behavior. Vibration of structural parts during prototype plants preoperational tests is measured using displacement gages, accelerometers, and strain transducers. The signals are recorded with F.M. magnetic tape records. On site and offsite signal analysis is done using both hybrid real time and digital techniques to determine the (approximate) frequency and phase content. In some structural components the spectral content of the signals include nearly discrete frequency or very narrow-band, usually due to excitation by the main coolant pumps and other components that reflect the response of the structure at a natural frequency to broad bands, mechanically and/or flow-induced excitation. Damping factors are also obtained from wave analyses. It is known from the theory of shells that the normal modes of a cylindrical shell can be expressed as sine and cosine combinations with indices m and n indicating the number of axial half waves and circumferential waves, respectively. The shape of each mode and the corresponding natural frequencies are functions of the numbers m and n. The general expression for the radial displacement of a simply supported shell is: 3.9-21
WBN M x w (x, , t) = m=1[A nm (t) cos n + Bnm (t) sin n ] sin n=0 L The shell vibration at a natural frequency depends on the boundary conditions at the ends. The effect of the ends is negligible for long shells or for higher order m modes, and long shells have the lowest frequency for n = 2 (elliptical mode). For short shells, the effects of the ends are more important, and the shell will tend to vibrate in modes corresponding to values of n > 2. In general, studies of the dynamic behavior of components follow two parallel procedures: 1) obtain frequencies and spring constants analytically, and 2) confirm these values from the results of the tests. Damping coefficients are established experimentally. Once these factors are established, the response can be computed analytically. In parallel, the responses of important reactor structures are measured during preoperational reactor tests and the frequencies and mode shapes of the structures are obtained. Theoretical and experimental studies have provided information on the added apparent mass of the water, which has the effect of decreasing the natural frequency of the component. For both cases, cross and parallel, the response is obtained after the forcing function and the damping of the system is determined. Pre- and post-hot functional inspection results, in the case of plants similar to prototypes, serve to confirm predictions that the internals are well behaved. Any gross motion or undue wear would be evident following the application of approximately 107 cycles of vibration expected during the test period. 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports and Core Support Structures 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits 3.9.3.1.1 Subsystems and Components Supplied by Westinghouse Design transients are presented in Section 5.2.1.5. For ASME Code Class 1 components, systems, and supports, loading conditions are presented in Section 5.2.1.10.1, and stress criteria are provided in Section 5.2.1.10.7. Additional information concerning methods of analysis is presented throughout Section 5.2.1.10. Results of analyses are documented in the stress reports that describe the system or components. For core support structures, design loading conditions are given in Section 4.2.2.3. Loading conditions are discussed in Section 4.2.2.4. 3.9-22
WBN In general, for reactor internals components and for core support structures, the criteria for acceptability, with regard to mechanical integrity analyses, are that adequate core cooling and core shutdown must be assured. This implies that the deformation of the reactor internals must be sufficiently small so that the geometry remains substantially intact. Consequently, the limitations established on the internals are concerned principally with the maximum allowable deflections and stability of the parts, in addition to a stress criterion to assure integrity of the components. For the LOCA plus the SSE condition, deflections of critical internal structures are limited. In a hypothesized downward vertical displacement of the internals, energy absorbing devices limit the displacement after contacting the vessel bottom head, ensuring that the geometry of the core remains intact. The following mechanical functional performance criteria apply:
- 1. Following the design basis accident, the functional criterion to be met for the reactor internals was that the plant shall be shutdown and cooled in an orderly fashion so that fuel cladding temperature was kept within specified limits. This criterion implies that the deformation of critical components must be kept sufficiently small to allow core cooling.
- 2. For large breaks, the reduction in water density greatly reduces the reactivity of the core, thereby shutting down the core whether the rods are tripped or not. The subsequent refilling of the core by the emergency core cooling system uses borated water to maintain the core in a subcritical state. Therefore, the main requirement was to assure effectiveness of the emergency core cooling system. Insertion of the control rods, although not needed, gives further assurance of ability to shut the plant down and keep it in a safe shutdown condition.
- 3. The inward upper barrel deflections are controlled to ensure no contacting of the nearest rod cluster control guide tube. The outward upper barrel deflections are controlled in order to maintain an adequate annulus for the coolant between the vessel inner diameter and core barrel outer diameter.
- 4. The rod cluster control guide tube deflections are limited to ensure operability of the control rods.
- 5. To ensure no column loading of rod cluster control guide tubes, the upper core plate deflection is limited.
Methods of analysis and testing for core support structures are discussed in Sections 3.9.1.3, 3.9.1.4.1, 3.9.2.3, 3.9.2.5, and 3.9.2.6. Stress limits and deformation criteria are given in Sections 4.2.2.4 and 4.2.2.5. 3.9-23
WBN 3.9.3.1.1.1 Plant Conditions and Design Loading Combinations For ASME Code Class 2 and 3 Components Supplied by Westinghouse Design pressure, temperature, and other loading conditions that provide the bases for design of fluid systems Code Class 2 and 3 components are presented in the sections which describe the systems. 3.9.3.1.1.2 Design Loading Combinations by Westinghouse The design loading combinations for ASME Code Class 2 and 3 equipment and supports are given in Table 3.9-1. The design loading combinations are categorized with respect to Normal, Upset, Emergency, and Faulted Conditions. Stress limits for each of the loading combinations are equipment oriented and are presented in Tables 3.9-2, 3.9-3, 3.9-4, and 3.9-6 for tanks, inactive pumps, valves, and active pumps, respectively. The definition of the stress equations and limits are in accordance with the ASME Code as follows: a.. For tanks and all other equipment, Section III of the ASME Code, 1971 Edition through Summer 1973 Addenda, and Code Cases 1607-1, 1635-1 and 1636-1 were utilized to establish stress limits for Normal, Upset, Emergency, and Faulted Conditions. In addition, Code Case 1657 was used for WBN Unit 2.
- b. Some equipment was provided in accordance with the Code Edition and Addenda in effect on the date of the contract.
For the actual numerical values of the allowables for specific equipment, the ASME Code Edition applicable to the time period of equipment procurement as specified on the procurement documents was used for the qualification. Active (Active components are those whose operability is relied upon to perform a safety function (as well as reactor shutdown function) during the transients or events considered in the respective operating condition categories) pumps and valves are discussed in Section 3.9.3.2. The equipment supports are designed in accordance with the requirements specified in Section 3.9.3.4. 3.9.3.1.1.3 Design Stress Limits By Westinghouse The design stress limits established for equipment are sufficiently low to assure that violation of the pressure retaining boundary will not occur. These limits, for each of the loading combinations, are equipment oriented and are presented in Tables 3.9-2 through 3.9-4, and 3.9-
- 6. See Section 3.9.3.1.1.2 for discussion of applicable code editions.
3.9-24
WBN 3.9.3.1.2 Subsystems and Components Analyzed or Specified by TVA A. ASME Code Class 1, 2, and 3 Piping. The analytic procedures and modeling of piping systems is discussed in Sections 3.7.3.8 and 3.7.3.3.* As discussed in Section 3.7.3.8.1 the TVA analysis effort has been categorized into two approaches: Rigorous and Alternate. The loading sources, conditions, and stress limits are described below for each category and the results are summarized for each.
- Generated reactor coolant loop response spectra curves and movements enveloping the Set B + Set C curves are used for the analysis or reanalysis of auxiliary piping systems attached to the reactor coolant loops. The ASME Code Case N-411 or Regulatory Guide 1.61 damping values can be used when Set B + Set C spectra are considered.
- 1. Loading Conditions, Stress Limits and Requirements for Rigorous Analysis
- a. The loading sources considered in the rigorous analysis of a piping system are defined in Table 3.9-7.
- b. The piping was analyzed to the requirements of applicable codes as defined in Section 3.7.3.8.1.
- c. The design load combinations are categorized with respect to Normal, Upset, Emergency, and Faulted Conditions. Class 1 piping was analyzed using the limits established in Table 3.9-8 for all applicable loading conditions. The pressurizer surge line was also evaluated for the thermal stratification and thermal striping in response to the NRC Bulletin 88-11.
Other rigorously analyzed piping meets the limits established in Table 3.9-9 for all applicable loading conditions.
- d. Consideration was given to the sequence of events in establishing which load sources are taken as acting concurrently.
- e. Equipment nozzle loads are within vendor and/or TVA allowable values.
This ensures that functionality and 'Active' equipment operability requirements are met.
- f. All equipment (i.e., valves, pumps, bellows, flanges, strainers, etc.) was checked to ensure compliance with vendor limitations.
- g. The pipe/valve interface at each active valve was evaluated and the pipe stresses are limited to the levels indicated in Table 3.9-10 unless higher limits are technically justified on a case-by-case basis.
3.9-25
WBN
- h. Documentation of rupture stress was provided for the locations in the system being analyzed where the stress exceeds the limits for which pipe rupture postulation was required (See Section 3.6). The tabulation identifies the point and tabulates the stress for each point exceeding the limits.
- i. Valves with extended operators or structures (including handwheels) meet the dynamic plus gravitational acceleration limits of 3g along the stem axis and 3g (vectorial summation) in the plane perpendicular to the valve stem axis. For 1-inch and smaller valves with handwheels, the dynamic plus gravitational acceleration limit is 3g in each of the three global (or local) directions. These limits apply to any valve orientation and must be maintained during piping analysis.
For steel body check valves (which have no external operators or structures) the limit for dynamic plus gravitational acceleration was 10g (vectorial summation of all three orthogonal directions). The valves as a minimum are qualified to the acceleration limits specified above. Higher accelerations are approved based on case-by-case technical justification.
- j. Excessive pipe deformation was avoided.
- k. Welded attachment loads and stresses for TVA Class 1 piping were evaluated in accordance with ASME Code Cases N-122 and N-391.
For Class 2 and 3 piping, loads and stresses from welded attachments were evaluated in accordance with ASME Code Cases N-318 and N-392. Special cases of other welded attachments were evaluated by detailed finite element analysis or other applicable methods to assure that ASME Code stress allowables were met. The attachment welds are full penetration, partial penetration, or fillet welded as detailed on the support drawings. Attachments are used generally on all piping systems, and locations can be determined from the support drawings.
- 2. Loading Conditions and Stress Limits for Alternate Analysis
- a. The scope of the alternate analysis application is generally limited to systems having the following load sources: self-weight, internal pressure, seismic event, end point displacement, and limited thermal expansion.
(Other load sources may be considered for special cases.) 3.9-26
WBN
- b. The design load combinations are categorized with respect to Normal, Upset, Emergency, and Faulted Conditions. The criteria are developed to meet the stress limits given in Table 3.9-9 considering the applicable load sources.
- c. The general limitations imposed on the piping by the application of the Alternate Analysis method are discussed in Section 3.7.3.8.3. For ASME Category I piping designed by alternate analysis, the same levels of valve acceleration and interface/nozzle load requirements of Section 3.9.3.1.2.A shall be maintained. Non-ASME, Category I(L) piping designed by alternate analysis is described in Sections 3.7.3.8.3 and 3.2.1.
- 3. Considerations for the Faulted Condition Tables 3.9-8 through 3.9-10 identify the load sources and allowed stresses associated with the faulted condition. The stress limits used are those limits established in ASME Section III for the faulted condition.
The feedwater system inside containment, from the check valves to the steam generators including the piping components are evaluated for pressure boundary integrity to withstand the postulated water hammer event due to the feedwater check valve slam following pipe rupture at the main header (Turbine Building) using the ASME Section III Appendix F (1980 Edition through Winter 1982 Addenda) rules and limits. The four main feedwater check valves were evaluated for structural integrity following the feedwater pipe rupture. Energy equivalence methods, in conjunction with nonlinear finite element and linear hand analyses, were used. The evaluations demonstrated that deformations in three of the four valves are within acceptable strain levels following the slam. With the assumption that the fourth valve is not functional, the transient effects of the resulting one steam generator blowdown are bounded by the "Major Rupture of a Main Feedwater Pipe inside containment" per Section 15.4.2.2. Note that during the rigorous analysis phase of most piping systems, the postulated break locations are unknown and the jet impingement loads are unavailable and thus not included in the evaluation of the faulted condition. However, where it was determined by the guidelines of Section 3.6 that jet impingement must be evaluated, the effect of the loads on pipe stress was evaluated during the pipe rupture analysis. 3.9-27
WBN
- 4. Summary of Results - Rigorous Analysis of Class 1 and Class 2/3 Piping Performed by TVA The results of the piping system analyses performed in accordance with the above paragraphs are presented and consolidated in calculations with the following documentation:
- a. Certification Report for ASME Code Class 1 Analyses.
- b. Owner's review for ASME Code Class 1 Analyses.
- c. Statement of Compliance with code requirements for ASME Code Class 2/3 Analyses.
- d. Problem revision status form - for maintaining the traceability of revision performed on analysis, and correlating various forms affected by each revision.
- e. Piping input data for recording all physical data used in the analysis.
- f. Table of system operating modes - for identifying the various thermal conditions required and included in analysis.
- g. Stress summary - for summarizing the maximum stresses for various loading combinations.
- h. Equipment nozzle load qualification to demonstrate satisfaction of limits.
- i. Valve acceleration qualification to demonstrate satisfaction of limits.
- j. Summary of loads and movements at pipe supports.
A record copy of these problem calculations is maintained at TVA and is available for review upon request. B. Category I ASME Code Class 2 and 3 Mechanical Equipment
- 1. Plant Conditions and Design Loading Combinations Design pressure, temperature, and other loading conditions that provide the bases for design of fluid systems Code Class 2 and 3 components are presented in the sections which describe the systems.
3.9-28
WBN
- 2. Design Loading Combinations The design loading combinations for ASME Code Class 2 and 3 equipment and supports are given in Table 3.9-13B. The design loading combinations are categorized with respect to Normal, Upset, Emergency and Faulted Conditions.
Stress limits for each of the loading combinations are equipment oriented and are presented in Tables 3.9-14, 3.9-15 and 3.9-16 for tanks, inactive* pumps, and inactive* valves respectively. The definition of the stress equations and limits are in accordance with the ASME Code as follows:
- Inactive components are those whose operability are not relied upon to perform a safety function during the transients or events considered in the respective operating condition category.
- a. For tanks and all other equipment, Section III of the ASME Code, 1971 Edition through Summer 1973 Addenda, and Code Cases 1607-1, 1635-1 and 1636-1 were utilized to establish stress limits for Normal, Upset, Emergency, and Faulted Conditions. In addition, Code Case 1657 was used for WBN Unit 2.
- b. Some equipment was provided in accordance with the Code Edition and Addenda in effect on the date of the award of contract.
For the actual numerical values of the allowables for specific equipment, the ASME Code Edition applicable to the time period of equipment procurement as specified on the procurement documents was used for the qualification. Active pumps and valves are discussed in Section 3.9.3.2.1. The vendor supplied equipment/component supports stress levels are limited to the allowable stress of AISC or ASME Section III subsection NF or other comparable stress limits as delineated in the applicable design specification. Section 3.8.4 describes the allowable stresses used for TVA-designed equipment/component supports. The design stress limits established for the components are sufficiently low to assure that violation of the pressure retaining boundary will not occur. These limits, for each of the loading combinations, are component oriented and are presented in Tables 3.9-14 through 3.9-16. 3.9.3.2 Pumps and Valve Operability Assurance 3.9.3.2.1 Active ASME Class 1, 2, & 3 Pumps and Valves Active components are those whose operability is relied upon to perform a safety function (as well as reactor shutdown function) during the transients or events considered in the respective operating condition categories. 3.9-29
WBN The list of active valves for primary fluid (i.e., water and steam containing components) systems in the Westinghouse scope of supply is presented in Table 3.9-17. The list of active pumps supplied by Westinghouse is presented in Table 3.9-28. The list of pumps and valves for fluid systems within TVA scope of supply are presented in Tables 3.9-25 and 3.9-27. Only ASME Section III pumps and valves that were purchased after September 1, 1974, were considered to be within the scope of WBN compliance with Regulatory Guide 1.48. These pumps and valves meet the special design requirements verifying operability as specified in Regulatory Guide 1.48. The remaining components in Tables 3.9-17, 3.9-25, 3.9-27, and 3.9-28 meet the appropriate qualification requirements in accordance with the guidelines of IEEE 344-1971 and consistent with the ASME Code applicable at the time of the contract date for procuring the component. These qualifications provide an adequate level of operability assurance for all active pumps and valves. The following rules were used to identify active pumps and valves:
- 1. Only UFSAR Chapter 15 Design Basis Events (DBE's) were assumed. These DBE's were studied to identify the active pumps and valves required to mitigate the DBE and place the plant in a safe shutdown condition.
- 2. Reactor Coolant Pressure Boundary (RCPB) - Valves that are a part of the RCPB (defined by 10 CFR Section 50.2) and require movement to isolate the RCS were identified as active.
- 3. Containment Isolation - Containment isolation valves that require movement to isolate the containment were identified as active.
- 4. Check Valves - Any check valve required to close or cycle when performing its system safety function was identified as active.
Any check valve that was only required to open in the performance of its system safety function was not identified as active. This position was justified by: (a) the free-swinging nature of the valves and (b) the normal stress over-design of the valve body.
- 5. Achieve and Maintain a Safe Shutdown Condition - The minimum redundant complement of equipment required to achieve and maintain safe shutdown was selected.
3.9.3.2.2 Operability Assurance 3.9.3.2.2.1 Westinghouse Scope of Supply Mechanical equipment classified as safety-related must be shown capable of performing its function during the life of the plant under postulated plant conditions. Equipment with faulted condition functional requirements include 'active' pumps and valves in fluid systems such as the residual heat removal system, safety injection system, and the containment spray system. Seismic analysis is presented in Section 3.7 and covers all safety-related mechanical equipment. 3.9-30
WBN Operability is assured by satisfying the requirements of the programs specified below. Additionally, equipment specifications include requirements for operability under the specified plant conditions and define appropriate acceptance criteria to ensure that the program requirements defined below are satisfied. Pump and Valve Qualification for Operability Program Active pumps are qualified for operability by first, being subjected to rigid tests both prior to installation in the plant and after installation in the plant. The in-shop tests include: 1) hydrostatic tests of pressure retaining parts to 150 percent times the design pressure times the ratio of material allowable stress at room temperature to the allowable stress value at the design temperature, 2) seal leakage tests, and 3) performance tests to determine total developed head, minimum and maximum head, net positive suction head (NPSH) requirements and other pump parameters. Also monitored during these operating tests are bearing temperatures and vibration levels. Bearing temperature limits are determined by the manufacturer, based on the bearing material, clearances, oil type, and rotational speed. These limits are approved by Westinghouse. Vibration limits are also determined by the manufacturer and are approved by Westinghouse. After the pump is installed in the plant, it undergoes the cold hydro-tests, hot functional test, and the required periodic inservice inspection and operation. These tests demonstrate that the pump will function as required during all normal operating conditions for the design life of the plant. In addition to these tests, the safety-related active pumps, are qualified for operability by assuring that the pump will start up, continue operating, and not be damaged during the faulted condition. The pump manufacturer was required to show by analysis correlated by test, prototype tests or existing documented data that the pump will perform its safety function when subjected to loads imposed by the maximum seismic accelerations and the maximum faulted nozzle loads. It was required that test or dynamic analysis be used to show that the lowest natural frequency of the pump is greater than 33 Hz. The pump, when having a natural frequency above 33 Hz, is considered essentially rigid. This frequency is sufficiently high to avoid problems with amplification between the component and structure for all seismic areas. A static shaft deflection analysis of the rotor was performed with the conservative SSE accelerations of 3g horizontal and 2g vertical acting simultaneously. The deflections determined from the static shaft analysis were compared to the allowable rotor clearances. The nature of seismic disturbances dictates that the maximum contact will be of short duration. If rubbing or impact is predicted, it is required that it be shown by prototype tests or existing documented data that the pump will not be damaged or cease to perform its design function. The effect of impacting on the operation of the pump was evaluated by analysis or by comparison of the impacting surfaces of the pump to similar surfaces of pumps which had been tested. In order to avoid damage during the faulted plant condition, the stresses caused by the combination of normal operating loads, SSE, dynamic system loads are limited to the limits indicated in Table 3.9-6. In addition, the pump casing stresses caused by the maximum faulted nozzle loads are limited to the stresses outlined in Table 3.9-6. 3.9-31
WBN The changes in operating rotor clearances caused by casing distortions due to these nozzle loads were considered. The maximum seismic nozzle loads combined with the loads imposed by the seismic accelerations were considered in an analysis of the pump supports. Furthermore, the calculated misalignment was shown to be less than that misalignment which could cause pump misoperation. The stresses in the supports are below those in Table 3.9-6; therefore, the support distortion is elastic and of short duration (equal to the duration of the seismic event). Performing these analyses with the conservative loads stated and with the restrictive stress limits of Table 3.9-6 as allowables, assure that critical parts of the pump are not damaged during the short duration of the faulted condition and that, therefore, the reliability of the pump for post-faulted condition operation is not impaired by the seismic event. If the natural frequency was found to be below 33 Hz, an analysis was performed to determine the amplified input accelerations necessary to perform the static analysis. The adjusted accelerations were determined using the same conservatisms contained in the 3g horizontal and 2g vertical accelerations used for 'rigid' structures. The static analysis was performed using the adjusted accelerations; the stress limits stated in Table 3.9-6 were satisfied. To complete the seismic qualification procedures, the pump motor was qualified for operation during the maximum seismic event. Any auxiliary equipment identified to be vital to the operation of the pump or pump motor, and which is not proven adequate for operation by the pump or motor qualifications, was separately qualified by meeting the requirements of IEEE Standard 344-1971 or -1975, as applicable, with the additional requirements and justifications outlined in this section. The program above gives the required assurance that the safety-related pump/motor assemblies will not be damaged and will continue operating under SSE loadings, and, therefore, will perform their intended functions. These proposed requirements take into account the complex characteristics of the pump and are sufficient to demonstrate and assure the seismic operability of the active pumps. Since the pump is not damaged during the faulted condition, the functional ability of active pumps after the faulted condition is assured since only normal operating loads and steady state nozzle loads exist. Since it is demonstrated that the pumps would not be damaged during the faulted condition, the post-faulted condition operating loads will be identical to the normal plant operating loads. This is assured by requiring that the imposed nozzle loads (steady-state loads) for normal conditions and post-faulted conditions are limited by the magnitudes of the normal condition nozzle loads. The post-faulted condition ability of the pumps to function under these applied loads is proven during the normal operating plant conditions for active pumps. Safety-related active valves must perform their mechanical motion at times of an accident. Assurance is supplied that these valves will operate during a seismic event. Tests and analyses were conducted to qualify active valves. 3.9-32
WBN The safety-related active valves were subjected to a series of stringent tests prior to service and during the plant life. Prior to installation, the following tests were performed: shell hydrostatic test to ASME Section III requirements, backseat and main seat leakage tests, disc hydrostatic test, and operational tests to verify that the valve will open and close. For the active valves qualification of electric motor operators for the environmental conditions (i.e., aging, radiation, accident environment simulation, etc.) refer to Section 3.11 and Regulatory Guide 1.73. Cold hydro tests, hot functional qualifications tests, periodic inservice inspections, and periodic inservice operations are performed in-situ to verify and assure the functional ability of the valve. These tests guarantee reliability of the valve for the design life of the plant. The valves are constructed in accordance with the ASME Boiler and Pressure Vessel Code, Section III. On all active valves, an analysis of the extended structure was performed for static equivalent seismic SSE loads applied at the center of gravity of the extended structure. The stress limits allowed in these analyses show structural integrity. The limits used for active Class 2 and 3 valves are shown in Table 3.9-4. In addition to these tests and analyses, a representative electric motor operated valve was tested for verification of operability during a simulated plant faulted condition event by demonstrating operational capabilities within the specified limits. The testing procedures are described below. The valve was mounted in a manner which represents typical valve installations. The valve included operator and limit switches if such are normally attached to the valve in service. The faulted condition nozzle loads were considered in the test in either of two ways: 1) loads equivalent to the faulted condition nozzle loads were limited such that the operability of the valve was not affected. The operability of the valve during a faulted condition was demonstrated by satisfying the following criteria:
- 1. All the active valves were designed to have a first natural frequency which is greater than 33 Hz, if it was practical to do so. If the lowest natural frequency of an active valve was less than 33 Hz, then the valve's mathematical model was included in the piping dynamic analysis, so as to assure the calculated valve acceleration does not exceed the values used in the static tests of the manufacturer's qualification program and to reflect the proper valve dynamic behavior.
- 2. The actuator and yoke of the representative motor operated valve system was statically deflected using an equivalent static load that simulates those conditions applied to the valve under faulted condition accelerations applied at the center of gravity of the operator alone in the direction of the weakest axis of the yoke. The design pressure of the valve was simultaneously applied to the valve during the static deflection tests.
- 3. The valve was cycled while in the deflected position. The time required to open or close the valve in the defected position was compared to similar data taken in the undeflected condition to evaluate the significance of any change.
The accelerations used for the static valve qualification are 3g horizontal and 2g vertical with the valve yoke axis vertical. The piping designer maintained the operator accelerations to these levels unless higher limits were technically justified on a case-by-case basis. 3.9-33
WBN The testing was conducted on the valves with extended structures which are most affected by acceleration, according to mass, length and cross-section of extended structures. Valve sizes which cover the range of sizes in service were qualified by the tests and the results are used to qualify all valves within the intermediate range of sizes. Valves which are safety-related but can be classified as not having an extended structure, such as check valves and safety valves, were considered separately. Check valves are characteristically simple in design and their operation will not be affected by seismic accelerations or the applied nozzle loads. The check valve design is compact and there are no extended structures or masses whose motion could cause distortions which could restrict operation of the valve. The nozzle loads due to seismic excitation will not affect the functional ability of the valve since the valve disc is typically designed to be isolated from the body wall. The clearance supplied by the design around the disc will prevent the disc from becoming bound or restricted due to any body distortions caused by nozzle loads. Therefore, the design of these valves is such that once the structural integrity of the valve is assured using standard design or analysis methods, the ability of the valve to operate is assured by the design features. In addition to these design considerations, the valve was subjected to the following tests and analysis: 1) in-shop hydrostatic test, 2) in-shop seat leakage test, and 3) periodic in-situ valve exercising and inspection to assure the functional ability of the valve. The pressurizer safety valves were qualified by the following procedures (these valves are also subjected to tests and analysis similar to check valves): stress and deformation analyses of critical items which may affect operability for faulted condition loads, in-shop hydrostatic and seat leakage tests, and periodic in-situ valve inspection. In addition to these tests, a static load equivalent to that applied by the faulted condition was applied at the top of the bonnet and the pressure will be increased until the valve mechanism actuates. Successful actuation within the design requirements of the valve assured its overpressurization safety capabilities during a seismic event. Using these methods, active valves were qualified for operability during a faulted event. These methods conservatively simulate the seismic event and assure that the active valves will perform their safety-related function when necessary. The above testing program for valves is conservative. Alternate valve operability testing, such as dynamic vibration testing, was allowed if it was shown to adequately assure the faulted condition functional ability of the valve system. 3.9-34
WBN Pump Motor and Valve Operator Qualification Active pump motors (and vital pump appurtenances) and active valve electric motor operators (and limit switches and pilot solenoid valves), were seismically qualified by meeting the requirements of IEEE Standard 344-1971 or 1975, as applicable. If the testing option was chosen, sine-beat testing was justified. This justification was provided by satisfying one or more of the following requirements to demonstrate that multi-frequency response is negligible or the sine-beat input is of sufficient magnitude to conservatively account for this effect.
- 1. The equipment response is basically due to one mode.
- 2. The sine-beat response spectra envelopes the floor response spectra in the region of significant response.
- 3. The floor response spectra consists of one dominant mode and has a peak at this frequency.
If the degree of coupling in the equipment is small, then single axis testing may have been justified. Multi-axis testing was required if there is considerable cross coupling; however, if the degree of coupling can be determined, then single axis testing will be used with the input sufficiently increased to include the effect of coupling on the response of the equipment. Seismic qualification by analysis alone, or by a combination of analysis and testing, has been used when justified. The analysis program can be justified by: 1) demonstrating that equipment being qualified is amendable to analysis, and 2) that the analysis be correlated with test or be performed using standard analysis techniques. 3.9.3.2.2.2 TVA Scope of Supply TVA used the following criteria to prescribe a suitable program to assure the functional adequacy of active Category I fluid system components (pumps and valves) under combined loading conditions. These criteria supplement or amend previously stated requirements for fluid system components in those cases where fluid system components are judged to be active (i.e., if they perform a required mechanical motion during the course of accomplishing a safety function). These criteria assure that all active seismic Category I fluid system components will maintain structural integrity and perform their safety functions under loadings, including seismic, associated with normal, upset, and faulted conditions. These criteria are similar to the accepted response to NRC Position for the TVA's Bellefonte Nuclear Plant units 1 and 2 concerning compliance with the requirements of Regulatory Guide 1.48. The exception is that the seismic qualification for Watts Bar is for a 2-dimensional earthquake, while for Bellefonte it is for a 3-dimensional earthquake. 3.9-35
WBN 3.9.3.2.3 Criteria For Assuring Functional Adequacy of Active Seismic Category I Fluid System Components (Pumps and Valves) and Associated Essential Auxiliary Equipment
- 1. The seismic design adequacy of Category I electrical power and control equipment and instrumentation directly associated with the active Category I pumps and valves is assured by seismically qualifying the components by analysis and/or testing in accordance with the requirements of IEEE Standard 344 (for applicable edition, refer to Section 3.9.2.2).
- 2. When either analysis or testing is used to demonstrate the seismic design adequacy of Category I components, the characteristics of the required input motion is specified by either response spectra, power spectral density function or time history data derived from the structure or system seismic analysis. When the testing method is used, random vibration input motion shall be used, but single frequency input, such as sine beats, may be used provided that:
- a. The characteristics of the required input motion indicate that the motion is dominated by one frequency.
- b. The anticipated response of equipment is adequately represented by one mode.
- c. The input has sufficient intensity and duration to excite all modes to the required magnitude, such that the testing response spectra will envelop the corresponding response spectra of the individual modes.
For equipment with more than one dominant frequency and for equipment supported near the base of the structure where some random components of the earthquake may remain, single frequency testing may still be applicable provided that the input has sufficient intensity and duration to excite all modes to the required magnitude, such that the testing response spectra will envelop the corresponding response spectra of the individual modes. When equipment responses along one direction are sensitive to the vibration frequencies along another perpendicular direction, in the case of single frequency testing, the time phasing of the inputs in the vertical and horizontal directions is such that a purely rectilinear resultant output is avoided. In both the testing and analysis procedure, the possible amplified design loads for vendor supplied equipment is considered as follows:
- a. If supports are tested, they were tested with the actual components mounted and operating or if the components are inoperative during the support test, the response at the equipment mounting locations were monitored and components were tested separately and the actual input to the equipment was more conservative in amplitude and frequency content than the monitored responses.
3.9-36
WBN
- b. The support analysis includes the component loads. Seismic restraints were used as applicable with their adequacy verified by either testing or analysis as described above.
- 3. Active Category I pumps were subjected to tests both prior to installation in the plant and after installation in the plant. The in-shop tests include (a) hydrostatic tests of pressure-retaining parts, (b) seal leakage tests, and (c) performance tests, while the pump is operated with flow, to determine total developed head, minimum and maximum head, net positive suction head (NPSH) requirements and other pump/motor parameters. Bearing temperatures and vibration levels were monitored during these operating tests. Both were shown to be below appropriate limits specified to the manufacturer for design of the pump. After the pump was installed in the plant, it was subject to cold hydro tests, or operational tests, hot functional tests, and the required periodic in-service inspection and operation.
- 4. Active Category I pumps were analyzed to show that the pump will operate normally when subjected to the maximum seismic accelerations and maximum seismic nozzle loads. Tests or dynamic analysis show that the lowest natural frequency of the pump is above 33 Hz, and thus considered essentially rigid. A static shaft deflection analysis of the rotor was performed with the conservative seismic accelerations of 1.5 times the applicable plant floor response spectra. The deflections determined from the static shaft analysis were compared to the allowable rotor clearances. Stresses caused by the combination of normal operating loads, seismic, and dynamic system loads were limited to the material elastic limit, as indicated in Table 3.9-18. The primary membrane stress (Pm) for the faulted condition loads were limited to 1.2Sh, or approximately 0.75 S (S =
yield stress). The primary membrane stress plus the primary bending stress (Pb) was limited to 1.8Sh, or approximately 1.l S. In addition, the pump nozzle stresses caused by the maximum seismic nozzle loads were limited to stresses outlined in Table 3.9-18. The maximum seismic nozzle loads were considered in an analysis of the pump supports to assure that a system misalignment cannot occur. If the natural frequency is found to be below 33 Hz, then analyses are performed to determine the amplified input accelerations necessary to perform the static analysis. The adjusted accelerations were determined using the same conservatism contained in the accelerations used for "rigid" structures. The static analysis was performed using the adjusted accelerations; the stress limits stated in Table 3.9-18 were satisfied. 3.9-37
WBN
- 5. Each type of active Category I pump motor is independently qualified for operating during the maximum seismic event. Any appurtenances which are identified to be vital to the operation of the pump or pump motor and which are not qualified for operation during the pump analysis or motor qualifications, shall also be separately qualified for operation at the accelerations each would see at its mounting. The pump motor and vital appurtenances are qualified by meeting the requirements of IEEE Standard 344-1971 or 1975 edition, depending on the procurement date (see Section 3.9.2.2.). If the testing option was chosen, sine-beat testing was justified by satisfying one or more of the following requirements to demonstrate that multi-frequency response is negligible or the sine-beat input is of sufficient magnitude to conservatively account for this effect.
- a. The equipment response is basically due to one mode.
- b. The sine-beat response spectra envelops the floor response spectra in the region of significant response.
- c. The floor response spectra consist of one dominant mode and have a peak at this frequency. The degree of mass or stiffness coupling in the equipment will, in general, determine if a single or multi-axis test is required. Multi-axis testing was required if there is considerable cross coupling. If coupling is very light, then single axis testing is justified; or, if the degree of coupling can be determined, then single axis testing was used with the input sufficiently increased to include the effect of coupling on the response of the equipment.
- 6. The post-faulted condition operating loads for active Category I pumps was considered identical to the normal plant operating loads. This was assured by requiring that the imposed nozzle loads (steady-state loads) for normal conditions and post-faulted conditions be limited by the magnitudes of the normal condition nozzle loads. Thus, the post-faulted condition ability of the pumps to function under these applied loads was proven during the normal operating plant conditions.
- 7. Active Category I valves, except check valves, were subjected to a series of stringent tests prior to installation and after installation in the plant. Prior to installation, the following tests were performed: (a) shell hydrostatic test, (b) backseat and main seat leakage tests, (c) disc hydrostatic test, (d) functional tests to verify that the valve will operate within the specified time limits when subjected to the design differential pressure prior to operating, and (e) operability qualification of motor and air operator control valves for the conditions over their installed life (i.e., aging, radiation, accident environment simulation, etc.) in accordance with the requirements of IEEE Standard 382 (see Section 3.11). Cold hydro qualification tests, preoperational tests, hot functional qualification tests, periodic inservice inspections, and periodic inservice operation were performed after installation to verify and assure functional ability of the valves. To the extent practicable, functional tests are performed after installation to verify that the valve will open and/or close in a time consistent with required safety functions.
3.9-38
WBN
- 8. Active Category I valves are designed using either stress analyses or the pressure containing minimum wall thickness requirements. An analysis of the extended structure is performed for static equivalent seismic loads applied at the center of gravity of the extended structure. The maximum stress limits allowed in these analyses confirms structural integrity and were the limits developed and accepted by the ASME for the particular ASME class of valve analyzed. The stress limits used for active Class 2 and 3 valves are given in Table 3.9-19. Class 1 valves were designed according to the rules of the ASME Boiler and Pressure Vessel Code, Section III, NB-3500.
- 9. Representative active Category I valves of each design type with overhanging structures (i.e., motor or pneumatic operator) were tested for verification of operability during a simulated seismic event by demonstrating operational capabilities within specified limits.
The testing is conducted on a representative number of valves. Valves from each of the primary safety-related design types (e.g., motor- operated gate valves) were tested. Valve sizes which cover the range of sizes in service were qualified by the tests and the results were used to qualify all valves within the intermediate range of sizes. Stress and deformation analyses are used to support the interpolation. The valve was mounted in a manner which is conservatively representative of a typical plant installation. The valve includes the operator and all appurtenances normally attached to the valve in service. The operability of the valve during a seismic event was verified by satisfying the following requirements:
- a. All active valves are designed to have a first natural frequency which is greater than 33 Hz if practical to do so. This may be shown by suitable test or analysis.
If the lowest natural frequency of an active valve is less than 33 Hz, the valve's mathematical model is included in the piping dynamic analysis. This assures the calculated valve acceleration does not exceed the values used in the static tests of the manufacturers qualification program and reflects the proper valve dynamic behavior.
- b. The actuator and yoke of the valve system were statically loaded an amount greater than that determined by an analysis as representing the applicable seismic accelerations applied at the center of gravity of the operator alone in the direction of the weakest axis of the yoke. The design pressure of the valve is simultaneously applied to the valve during the static deflection tests.
- c. The valve was operated while in the deflected position, i.e., from the normal operating mode to the faulted operating mode. The valve performed its safety-related function within the specified operating time limits.
- d. Motor operators and other electrical appurtenances necessary for operation were qualified as operable during the seismic event by analysis and/or testing in accordance with the requirements of IEEE Standard 344 (refer to Section 3.9.2.2 for the applicable edition).
3.9-39
WBN The accelerations used for the static valve qualification are 3.0 g horizontal and 2.0 g vertical with the valve yoke axis vertical. The piping designer shall maintain the motor operator accelerations to equivalent levels. If the valve accelerations exceed these levels, an evaluation of the valve is performed to document acceptability on a case-by-case basis. If the frequency of the valve, by test or analysis, was less than 33 Hz, a dynamic analysis of the valve was performed to determine the equivalent acceleration to be applied during the static test. The analysis provided the amplification of the input acceleration considering the natural frequency of the valve and piping along with the frequency content of the applicable plant floor response spectra. The adjusted accelerations were determined using the same conservatism contained in the accelerations used for "rigid" valves. The adjusted accelerations were then used in the static analysis and the valve operability was assured by the methods outlined in steps (b), (c), and (d) above using the modified acceleration input.
- 10. The design of each active Category I check valve was such that once the structural integrity of the valve was assured using standard design or analysis methods, the ability of the valve to operate was assured by the design features. In addition to design considerations, each active check valve undergoes:
- a. Stress analysis including the applicable seismic loads
- b. In-shop hydrostatic tests for parts that could affect the operability of the valve,
- c. In-shop seat leakage tests, and
- d. Preoperational and periodic in-situ testing and inspection to assure functional ability of the valves.
- 11. The design of the pressurizer safety valve (Category I) was such that once the structural integrity of the valves was assured using standard design or analysis methods, the ability of the valve to operate was assured by the design features. In addition to design considerations, the pressurizer safety valve was subjected to:
- a. Stress and/or deformation analyses for parts that could affect the operability of the valv}}