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, EQUIPMENT AND SYSTEMS
, EQUIPMENT AND SYSTEMS


S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 1  3.1 CONFORMANCE TO NRC GE NERAL DESIGN CRITERIA This section discusses the extent to which the design criteria for plant structures, systems and components important to safety meet the NRC Ge neral Design Criteria for Nuclear Power Plants, specified in Appendix A to 10 CFR Part 50. For each criterion, a summary is provided which discusses how the principal design features meet the criterion and identifies any exceptions that are taken. In the discussion of each criterion, th e chapters or sections of this Updated FSAR where more detailed information is presented are referenced to demonstrate compliance with the criterion.
S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 1  3.1 CONFORMANCE TO NRC GE NERAL DESIGN CRITERIA This section discusses the extent to which the design criteria for plant structures, systems and components important to safety meet the NRC Ge neral Design Criteria for Nuclear Power Plants, specified in Appendix A to 10 CFR Part 50. For each criterion, a summary is provided which discusses how the principal design features meet the criterion and identifies any exceptions that are taken. In the discussion of each criterion, th e chapters or sections of this Updated FSAR where more detailed information is presented are referenced to demonstrate compliance with the criterion.  
3.1.1 Overall Requirements 3.1.1.1 Criterion 1 - Quality Standards and Records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keep ing with th e required safety function. A quality assurance program sh all be established and implemented to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropr iate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licen see throughout the life of the unit.
 
====3.1.1 Overall====
Requirements 3.1.1.1 Criterion 1 - Quality Standards and Records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keep ing with th e required safety function. A quality assurance program sh all be established and implemented to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropr iate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licen see throughout the life of the unit.
RESPONSE The structures, systems and components of the facility are classified according to their importance in the prevention and mitigation of accidents, using the classification system developed by the American Nuclear Society. Each safety-related component is given a safety class designation. The codes, standards and quality controls applicable to each type of safety-related component and its safety class designation are identified in Chapter 3. Where applicable, design, fabrication, er ection and testing are in accordance with the codes required in 10 CFR 50.55a.
RESPONSE The structures, systems and components of the facility are classified according to their importance in the prevention and mitigation of accidents, using the classification system developed by the American Nuclear Society. Each safety-related component is given a safety class designation. The codes, standards and quality controls applicable to each type of safety-related component and its safety class designation are identified in Chapter 3. Where applicable, design, fabrication, er ection and testing are in accordance with the codes required in 10 CFR 50.55a.
Recognized codes and standards, when used, are identified and evaluated to assure their applicability, adequacy and sufficiency in keeping with the requi red safety function.
Recognized codes and standards, when used, are identified and evaluated to assure their applicability, adequacy and sufficiency in keeping with the requi red safety function.
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S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 37  3.1.6.5 Criterion 64 - Monitoring Radioactivity Releases Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released for normal operations, including anticipated operational occurrences, and from postulated accidents.
S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 37  3.1.6.5 Criterion 64 - Monitoring Radioactivity Releases Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released for normal operations, including anticipated operational occurrences, and from postulated accidents.
RESPONSE The containment atmosphere is continually monitored during normal and transient station operations, using the containment air particulate and radio-gas radiation monitors. In accident conditions, samples of the containment atmosphere provide data on existi ng airborne radioactive concentrations within the containment. Radioactivity levels contained within the facility gaseous and liquid effluent discharge paths and in the station environs are continually monitored during normal and accident conditions by the station radiation monitoring as described in Chapter 11 and 12. In addition to the installed detection system, an environment radiation surveillance program monitors the exposure pathways to Man within a 5-mile radius from the station site.
RESPONSE The containment atmosphere is continually monitored during normal and transient station operations, using the containment air particulate and radio-gas radiation monitors. In accident conditions, samples of the containment atmosphere provide data on existi ng airborne radioactive concentrations within the containment. Radioactivity levels contained within the facility gaseous and liquid effluent discharge paths and in the station environs are continually monitored during normal and accident conditions by the station radiation monitoring as described in Chapter 11 and 12. In addition to the installed detection system, an environment radiation surveillance program monitors the exposure pathways to Man within a 5-mile radius from the station site.
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 1  3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 3.2.1 Seismic Classification Seabrook Station structures, systems and components important to safety, as well as their foundations and supports, have been designed to withstand the effects of an Operating Basis Earthquake (OBE) and a Safe Shutdown Earthquake (SSE), and are thus designated as seismic Category I. These plant features are those necessary to assure: a. The integrity of the reactor coolant pressure boundary, b. The capability to shut down the reactor and maintain it in a safe shutdown condition, or c. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comp arable to the guideline exposures of 10 CFR 100. Those structures, systems and components classified as seismic Category I are listed in Table 3.2-1 and Table 3.2-2. These structures, systems and components are designed to withstand seismic loads as discussed in Section 3.7.
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 1  3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS
 
====3.2.1 Seismic====
Classification Seabrook Station structures, systems and components important to safety, as well as their foundations and supports, have been designed to withstand the effects of an Operating Basis Earthquake (OBE) and a Safe Shutdown Earthquake (SSE), and are thus designated as seismic Category I. These plant features are those necessary to assure: a. The integrity of the reactor coolant pressure boundary, b. The capability to shut down the reactor and maintain it in a safe shutdown condition, or c. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comp arable to the guideline exposures of 10 CFR 100. Those structures, systems and components classified as seismic Category I are listed in Table 3.2-1 and Table 3.2-2. These structures, systems and components are designed to withstand seismic loads as discussed in Section 3.7.
Seismic Category I structures are sufficiently isolated or protected for non-seismic Category I structures to ensure that their integrity is maintained at all times. None of the plant structures is classified as partially seismic Category I, with the exception of the Circulating Water Pumphouse and the Waste Processing Building, which are described in Section 3.8. Several non-seismic Category I structures are designed against collapse onto seismic Category I structures due to SSE loadings. These are discussed in Subsection 3.7(B).2.
Seismic Category I structures are sufficiently isolated or protected for non-seismic Category I structures to ensure that their integrity is maintained at all times. None of the plant structures is classified as partially seismic Category I, with the exception of the Circulating Water Pumphouse and the Waste Processing Building, which are described in Section 3.8. Several non-seismic Category I structures are designed against collapse onto seismic Category I structures due to SSE loadings. These are discussed in Subsection 3.7(B).2.
Where only portions of systems are classified as seismic Category I, the boundaries of the seismic Category I portions of the system are shown on the ductwork, piping and instrumentation diagrams in the referenced sections of the Updated FSAR. The seismic Category I portion of these systems extends to the first restraint beyond th e isolation valves that isolate the part which is Category I from the nonseismic portion of the system. Equipment and components that are not classified as seismic Category I, and whose collapse or failure could result in the loss of safety function of seismic Category I structures, systems or components, are checked to confirm their structur al integrity against collapse or failure due to SSE loadings.
Where only portions of systems are classified as seismic Category I, the boundaries of the seismic Category I portions of the system are shown on the ductwork, piping and instrumentation diagrams in the referenced sections of the Updated FSAR. The seismic Category I portion of these systems extends to the first restraint beyond th e isolation valves that isolate the part which is Category I from the nonseismic portion of the system. Equipment and components that are not classified as seismic Category I, and whose collapse or failure could result in the loss of safety function of seismic Category I structures, systems or components, are checked to confirm their structur al integrity against collapse or failure due to SSE loadings.
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 2  Where the collapse on non-Category I components would adversely affect the performance of Category I systems or components, they have been supported to withstand seismic loadings or isolated from the Category I systems or compone nts by boundary restraints. The locations of the boundary restraints are shown on the pipe support location drawings and the pipe support detail drawings. Where the collapse of non-Category I HVAC system components or ductwork would damage the other Category I system or components in the vi cinity, they have been supported to withstand seismic loadings.
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 2  Where the collapse on non-Category I components would adversely affect the performance of Category I systems or components, they have been supported to withstand seismic loadings or isolated from the Category I systems or compone nts by boundary restraints. The locations of the boundary restraints are shown on the pipe support location drawings and the pipe support detail drawings. Where the collapse of non-Category I HVAC system components or ductwork would damage the other Category I system or components in the vi cinity, they have been supported to withstand seismic loadings.
The seismic classifications presented in Table 3.2-1 and Table 3.2-2 are consistent with the recommendations of NRC Regulatory Guide 1.29.
The seismic classifications presented in Table 3.2-1 and Table 3.2-2 are consistent with the recommendations of NRC Regulatory Guide 1.29.  
3.2.2 Quality Group Classification The system quality group classification applies to fluid systems, or portions of fluid systems, which are directly depended upon to: a. Prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary b. Permit shutdown of the reactor and maintenance in the safe shutdown condition c. Contain radioactive material. Fluid system components important to safety are classified in accordance with the ANSI N18.2a-1975, "Revision and Addendum to Nuclear Sa fety Criteria for the Design of Stationary Pressurized Water Reactor Plants."  This classification system is compatible with the requirements of NRC Regulatory Guide 1.26, and is submitted as an alternate acceptable method of meeting the intent of Regulatory Guide 1.26.
 
====3.2.2 Quality====
Group Classification The system quality group classification applies to fluid systems, or portions of fluid systems, which are directly depended upon to: a. Prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary b. Permit shutdown of the reactor and maintenance in the safe shutdown condition c. Contain radioactive material. Fluid system components important to safety are classified in accordance with the ANSI N18.2a-1975, "Revision and Addendum to Nuclear Sa fety Criteria for the Design of Stationary Pressurized Water Reactor Plants."  This classification system is compatible with the requirements of NRC Regulatory Guide 1.26, and is submitted as an alternate acceptable method of meeting the intent of Regulatory Guide 1.26.
3.2.2.1 Safety Class Definitions Components are classified as Safety Class 1, Safety Class 2, Safety Class 3, and nonnuclear-safety in accordance with their importance to nuclear safety. This importance, as established by the assigned safety class, is applied in the design, materials, manufacture or fabrication, assembly, erection, construction, and operation. A single system may have components in more than one safety class.
3.2.2.1 Safety Class Definitions Components are classified as Safety Class 1, Safety Class 2, Safety Class 3, and nonnuclear-safety in accordance with their importance to nuclear safety. This importance, as established by the assigned safety class, is applied in the design, materials, manufacture or fabrication, assembly, erection, construction, and operation. A single system may have components in more than one safety class.
The definitions of safety classes listed apply to fluid pressure boundary components and the reactor containment. Supports that have a nuclear safety function are of the same safety class as the components that they support.
The definitions of safety classes listed apply to fluid pressure boundary components and the reactor containment. Supports that have a nuclear safety function are of the same safety class as the components that they support.
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C p = pressure coefficient (Table 4, Reference 1)
C p = pressure coefficient (Table 4, Reference 1)
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 2    The vertical profile of wind pressure on the Containment Enclosure Building is summarized in Table 3.3-1. The distributi on of effective pressure coefficients, (C p), for cylindrical and spherical structures is shown on Figure 3.3-1. b. ANSI - The effective velocity pressure vertical profile, including gust factors for structures (q F), parts of structures (q p), and internal pressure (q M) are taken from Tables 5, 6 and 12 (Exposure C) respectively of Reference 2. The effective velocity pressures are summarized in Ta ble 3.3-2. The average pressure acting on a structural element is as follows: p = q C p - q M C pi where:
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 2    The vertical profile of wind pressure on the Containment Enclosure Building is summarized in Table 3.3-1. The distributi on of effective pressure coefficients, (C p), for cylindrical and spherical structures is shown on Figure 3.3-1. b. ANSI - The effective velocity pressure vertical profile, including gust factors for structures (q F), parts of structures (q p), and internal pressure (q M) are taken from Tables 5, 6 and 12 (Exposure C) respectively of Reference 2. The effective velocity pressures are summarized in Ta ble 3.3-2. The average pressure acting on a structural element is as follows: p = q C p - q M C pi where:
p = average wind pressure (psf) q = q F or q p whichever is appropriate C p = external pressure coefficient q M = internal velocity pressure C pi = internal pressure coefficient Load combinations involving normal wind (discussed in Subsections 3.8.4.3 and 3.8.5.3) do not control the design of seismic Category I structures.
p = average wind pressure (psf) q = q F or q p whichever is appropriate C p = external pressure coefficient q M = internal velocity pressure C pi = internal pressure coefficient Load combinations involving normal wind (discussed in Subsections 3.8.4.3 and 3.8.5.3) do not control the design of seismic Category I structures.  
3.3.2 Tornado Loadings 3.3.2.1 Applicable Design Parameters
 
====3.3.2 Tornado====
Loadings 3.3.2.1 Applicable Design Parameters
: a. Tornado wind loads on seismic Category I structures are based on a single vortex tornado model which results in a maximum tangential velocity of 290 mph and a maximum translational velocity of 70 mph. The total maximum wind velocity is 360 mph assumed constant with respect to height. The pressure differential is 3 psi at a rate of 2 psi/second. See Subsection 2.3.1.2 for a discussion of the design basis tornado parameters.
: a. Tornado wind loads on seismic Category I structures are based on a single vortex tornado model which results in a maximum tangential velocity of 290 mph and a maximum translational velocity of 70 mph. The total maximum wind velocity is 360 mph assumed constant with respect to height. The pressure differential is 3 psi at a rate of 2 psi/second. See Subsection 2.3.1.2 for a discussion of the design basis tornado parameters.
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 3  b. Protection of equipment and systems locat ed in the "closed" structures listed in paragraph c. below, is accomplished by using protective barriers consisting of dampers complete with the necessary accessories and instrumentation. These dampers are provided at inlet openings located in the exterior walls of the structures which provide outdoor air for the plant HVAC systems. These dampers are also provided at the openings in the exterior walls of the structures, through which air is exhausted from the various plant HVAC systems to the outside atmosphere. The dampers automatically close to avoid damage which could be  
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 3  b. Protection of equipment and systems locat ed in the "closed" structures listed in paragraph c. below, is accomplished by using protective barriers consisting of dampers complete with the necessary accessories and instrumentation. These dampers are provided at inlet openings located in the exterior walls of the structures which provide outdoor air for the plant HVAC systems. These dampers are also provided at the openings in the exterior walls of the structures, through which air is exhausted from the various plant HVAC systems to the outside atmosphere. The dampers automatically close to avoid damage which could be  
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S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 7  3.3.3 References
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 7  3.3.3 References
: 1. ASCE Paper No. 3269, "Wind Forces on Structures," Transactions, Vol. 126, Part II, 1961, P1124. 2. ANSI A58.1-1972, "Building Code Requirements for Minimum Design Loads in Buildings and Others Structures," Am erican National Standards Institute, New York, 1972.
: 1. ASCE Paper No. 3269, "Wind Forces on Structures," Transactions, Vol. 126, Part II, 1961, P1124. 2. ANSI A58.1-1972, "Building Code Requirements for Minimum Design Loads in Buildings and Others Structures," Am erican National Standards Institute, New York, 1972.
SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Water Level (Flood) Design Revision 16 Section 3.4 Page 1  3.4  WATER LEVEL (FLOOD) DESIGN 3.4.1 Flood Protection 3.4.1.1  Flood Protection Measures for Seismic Category I Structures All safety
SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Water Level (Flood) Design Revision 16 Section 3.4 Page 1  3.4  WATER LEVEL (FLOOD) DESIGN
 
====3.4.1 Flood====
Protection 3.4.1.1  Flood Protection Measures for Seismic Category I Structures All safety
-related systems and components, as identified in Table 3.2-1 and Table 3.2-2, are protected against floods. These systems and components are protected by the structures which house them, and/or by being located above a maximum water level not exceeding 21 feet MSL, postulated to result during the combined PMH
-related systems and components, as identified in Table 3.2-1 and Table 3.2-2, are protected against floods. These systems and components are protected by the structures which house them, and/or by being located above a maximum water level not exceeding 21 feet MSL, postulated to result during the combined PMH
-SPS event.
-SPS event.
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-related structures, systems and components are provided with protection against floods, as described in the preceding paragraphs, it will not be necessary to bring the reactor to a cold shutdown for the most severe flood anticipated at the plant.
-related structures, systems and components are provided with protection against floods, as described in the preceding paragraphs, it will not be necessary to bring the reactor to a cold shutdown for the most severe flood anticipated at the plant.


SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Water Level (Flood) Design Revision 16 Section 3.4 Page 2  Protection against possible flooding from liquid carrying systems due to pipe rupture or fire protection activities is discussed in the following sections:  Section 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Subsection 9.3.3 Equipment and Floor Drainage System Subsection 9.5.1 Fire Protection System 3.4.1.2 Plant Dewatering System Seabrook Station was not originally designed with a dewatering system, because it was believed that the in
SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Water Level (Flood) Design Revision 16 Section 3.4 Page 2  Protection against possible flooding from liquid carrying systems due to pipe rupture or fire protection activities is discussed in the following sections:  Section 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Subsection
 
====9.3.3 Equipment====
and Floor Drainage System Subsection 9.5.1 Fire Protection System 3.4.1.2 Plant Dewatering System Seabrook Station was not originally designed with a dewatering system, because it was believed that the in
-leakage prevention methods described in 3.4.1.1 would be adequate to prevent water ingress. Over the years, it has become evident that the mitigation methods were not completely effective at preventing in
-leakage prevention methods described in 3.4.1.1 would be adequate to prevent water ingress. Over the years, it has become evident that the mitigation methods were not completely effective at preventing in
-leakage. The presence of groundwater in
-leakage. The presence of groundwater in
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-32' elevation. This pump discharges the ground water in the containment annulus to the roof drain system. The connection to the drain system is installed at 240 degrees azimuth of the containment annulus. Routine monitoring of this flow path is performed per station operating procedures.
-32' elevation. This pump discharges the ground water in the containment annulus to the roof drain system. The connection to the drain system is installed at 240 degrees azimuth of the containment annulus. Routine monitoring of this flow path is performed per station operating procedures.
A ground water collection tank and pump are installed in the "B" Electrical Tunnel, West stairwell at (-) 20'-0" elevation. The pump discharges the water to the turbine building roof drain system, which then flows to the storm drain system and out to circulating water for discharge via the outfall. Routine monitoring of this flow path is performed per station operating procedure. A 50.59 Applicability and Screening are performed to support this change.
A ground water collection tank and pump are installed in the "B" Electrical Tunnel, West stairwell at (-) 20'-0" elevation. The pump discharges the water to the turbine building roof drain system, which then flows to the storm drain system and out to circulating water for discharge via the outfall. Routine monitoring of this flow path is performed per station operating procedure. A 50.59 Applicability and Screening are performed to support this change.
3.4.2 Analytical and Test Procedures The methods and procedures by which the static and dynamic effects of the design basis flood conditions or design basis groundwater conditions identified in Section 2.4 are applied to safety-related structures, systems and components are described in Subsections 3.8.1, 3.8.4 and 3.8.5. In the design of seismic Category I structures, the groundwater is assumed to be at 20 feet MSL, the plant grade. As discussed in Subsection 3.4.1, the design basis flood could cause a depth of stillwater on the plant grade of not exceeding 1 foot, increasing the top of the hydrostatic head to elevation not exceeding 21 feet MSL. Dynamic effects of the design basis flood were considered, but found to be negligible. The maximum depth of stillwater is not exceeding one foot above plant grade, and the maximum wave runup in local regions is 1.8 feet above plant grade. Any dynamic effects produced by these occurrences were evaluated and found to be negligible and, due to the relatively large masses of the reinforced concrete structures, can be neglected.
 
====3.4.2 Analytical====
and Test Procedures The methods and procedures by which the static and dynamic effects of the design basis flood conditions or design basis groundwater conditions identified in Section 2.4 are applied to safety-related structures, systems and components are described in Subsections 3.8.1, 3.8.4 and 3.8.5. In the design of seismic Category I structures, the groundwater is assumed to be at 20 feet MSL, the plant grade. As discussed in Subsection 3.4.1, the design basis flood could cause a depth of stillwater on the plant grade of not exceeding 1 foot, increasing the top of the hydrostatic head to elevation not exceeding 21 feet MSL. Dynamic effects of the design basis flood were considered, but found to be negligible. The maximum depth of stillwater is not exceeding one foot above plant grade, and the maximum wave runup in local regions is 1.8 feet above plant grade. Any dynamic effects produced by these occurrences were evaluated and found to be negligible and, due to the relatively large masses of the reinforced concrete structures, can be neglected.
The coincident wind loadings used in the design of the structures are those associated with the Probable Maximum Hurricane (PMH), during which the design basis flood occurs. These wind loading are discussed in Subsection 3.3.1. The static and dynamic effects on foundations due to the design basis flood are considered in the design of all seismic Category I structures. Dynamic forces due to wave runup are negligible compared to the forces that are considered because of other design criteria.
The coincident wind loadings used in the design of the structures are those associated with the Probable Maximum Hurricane (PMH), during which the design basis flood occurs. These wind loading are discussed in Subsection 3.3.1. The static and dynamic effects on foundations due to the design basis flood are considered in the design of all seismic Category I structures. Dynamic forces due to wave runup are negligible compared to the forces that are considered because of other design criteria.
All seismic Category I structures are designed to prevent uplift or overturning, considering the effects of buoyant forces along with other forces (see Subsection 3.8.5).
All seismic Category I structures are designed to prevent uplift or overturning, considering the effects of buoyant forces along with other forces (see Subsection 3.8.5).
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A p = sectional pressure  = W/A lbs./sq ft W = weight of missile in lbs.
A p = sectional pressure  = W/A lbs./sq ft W = weight of missile in lbs.
A = cross section of missile in ft 2 V' = 215,000 V 1 log 2 10 V = velocity of missile in ft/sec  For barriers having thickness t < 3D, the penetration D' was calculated as: 2a'4e1DD'  where: a'  =  t/D  The minimum thickness required to prevent scabbing for steel missiles having solid circular cross section was determined from Figure 3.5-4. For other noncompressible missiles, design for no scabbing was based on t  3D, Reference 11.
A = cross section of missile in ft 2 V' = 215,000 V 1 log 2 10 V = velocity of missile in ft/sec  For barriers having thickness t < 3D, the penetration D' was calculated as: 2a'4e1DD'  where: a'  =  t/D  The minimum thickness required to prevent scabbing for steel missiles having solid circular cross section was determined from Figure 3.5-4. For other noncompressible missiles, design for no scabbing was based on t  3D, Reference 11.
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 30    Since the PSAR, experimental data ha s been generated by Sandia Laboratory for the Electric Power Research Institute for the specific missiles, velocities and targets applicable to these structures and barriers. The semi-analytical, semi-empirical modified National Defense Research Committee (NDRC) equations have been found to be more accurate than the modified Petry equations, Reference 12. The Modified NDRC formula for penetration is:
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 30    Since the PSAR, experimental data ha s been generated by Sandia Laboratory for the Electric Power Research Institute for the specific missiles, velocities and targets applicable to these structures and barriers. The semi-analytical, semi-empirical modified National Defense Research Committee (NDRC) equations have been found to be more accurate than the modified Petry equations, Reference 12. The Modified NDRC formula for penetration is:  
1.8 1000d VDWNK4x 2.0dxfor  where:  x = penetration depth in inches  K = c f'180/where f' c is the concrete design strength  N = missile shape factor  = 0.72 for a flat nose missile W = weight of missile in lbs.
 
===1.8 1000d===
VDWNK4x 2.0dxfor  where:  x = penetration depth in inches  K = c f'180/where f' c is the concrete design strength  N = missile shape factor  = 0.72 for a flat nose missile W = weight of missile in lbs.
d = diameter of missile in inches V = striking velocity of missile in ft/sec The minimum thickness to prevent scabbing is given by:
d = diameter of missile in inches V = striking velocity of missile in ft/sec The minimum thickness to prevent scabbing is given by:
t s/d  =  2.12 + 1.36 x/d for 0.65  x/d  11.75  where for pipes:  d = d e = effective diameter as though a solid cylinder S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 31  A factor of safety of 1.2 was applied to t
t s/d  =  2.12 + 1.36 x/d for 0.65  x/d  11.75  where for pipes:  d = d e = effective diameter as though a solid cylinder S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 31  A factor of safety of 1.2 was applied to t
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Page 24  3.7(N).5 References
Page 24  3.7(N).5 References
: 1. "Damping Values of Nuclear Plant Components," WCAP-7921-AR, May 1974. 2. Geninski, T. L. and Chiang, D., "Safety Analysis of the 17x17 Fuel Assembly for combined Seismic and Loss-of-Coolant Accident," WCAP-8236 (Proprietary), December 1973 and WCAP-8286 (Nonproprietary), December 1974. 3. Lin, C. W., "How to Lump the Masses - A Guide to the Piping Seismic Analysis," ASME paper 74-NE-7 presented at the Pr essure Vessels and Piping Conference, Miami, Florida, June 1974. 4. Biggs, J. M., Introduction to Structural Dynamics , McGraw-Hill, New York, 1964. 5. Thomas, T. H., et al., "Nuclear Reactors and Earthquakes," TID-7024, U.S. Atomic Energy Commission, Washington, D. C., August 1963. 6. Chan, S. P., Cox, H. L. and Benfield, W. A., "Transient Analysis of Forced Vibration of Complex Structural-Mechanical Systems," J. Royal Aeronautical Society," July 1962. 7. "Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March 1974 and WCAP-8288, Addendum 1 (Non-proprietary), April 1974.
: 1. "Damping Values of Nuclear Plant Components," WCAP-7921-AR, May 1974. 2. Geninski, T. L. and Chiang, D., "Safety Analysis of the 17x17 Fuel Assembly for combined Seismic and Loss-of-Coolant Accident," WCAP-8236 (Proprietary), December 1973 and WCAP-8286 (Nonproprietary), December 1974. 3. Lin, C. W., "How to Lump the Masses - A Guide to the Piping Seismic Analysis," ASME paper 74-NE-7 presented at the Pr essure Vessels and Piping Conference, Miami, Florida, June 1974. 4. Biggs, J. M., Introduction to Structural Dynamics , McGraw-Hill, New York, 1964. 5. Thomas, T. H., et al., "Nuclear Reactors and Earthquakes," TID-7024, U.S. Atomic Energy Commission, Washington, D. C., August 1963. 6. Chan, S. P., Cox, H. L. and Benfield, W. A., "Transient Analysis of Forced Vibration of Complex Structural-Mechanical Systems," J. Royal Aeronautical Society," July 1962. 7. "Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March 1974 and WCAP-8288, Addendum 1 (Non-proprietary), April 1974.
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 1  3.8 DESIGN OF CATEGORY I STRUCTURES 3.8.1 Concrete Containment The containment structure houses the major portion of a PWR Nuclear Steam Supply System (NSSS). During the operating life of the plant, it will also provide the following functions:
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 1  3.8 DESIGN OF CATEGORY I STRUCTURES
 
====3.8.1 Concrete====
Containment The containment structure houses the major portion of a PWR Nuclear Steam Supply System (NSSS). During the operating life of the plant, it will also provide the following functions:
: a. Limiting the leakage rate to the maximum allowable Type "A" test leakage rate, 0.15 percent by weight of the containment contained air mass per day at calculated peak pressure and associated temperature, resulting from any loss-of-coolant accident (LOCA) and other postulated accidents.
: a. Limiting the leakage rate to the maximum allowable Type "A" test leakage rate, 0.15 percent by weight of the containment contained air mass per day at calculated peak pressure and associated temperature, resulting from any loss-of-coolant accident (LOCA) and other postulated accidents.
: b. Providing continuing radiation shielding during normal plant operation in accordance with 10 CFR 20 and during accident conditions in accordance with 10 CFR 100. c. Protecting the reactor vessel and all other safety
: b. Providing continuing radiation shielding during normal plant operation in accordance with 10 CFR 20 and during accident conditions in accordance with 10 CFR 100. c. Protecting the reactor vessel and all other safety
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For the design of the liner, the load combinations in Table 3.8-1 are applicable with the exception that coefficients for all load cases are taken equal to 1.0.
For the design of the liner, the load combinations in Table 3.8-1 are applicable with the exception that coefficients for all load cases are taken equal to 1.0.
Steel components of the containment shell that are pressure
Steel components of the containment shell that are pressure
-resisting but unbacked by concrete are designed in accordance  with the ASME Code, Section III, Division 1, using loads described in Subsection 3.8.2. Earthquake effects are not assumed to occur simultaneously with flooding effects since the maximum flood is not associated with an earthquake.
-resisting but unbacked by concrete are designed in accordance  with the ASME Code, Section III, Division 1, using loads described in Subsection
 
====3.8.2. Earthquake====
effects are not assumed to occur simultaneously with flooding effects since the maximum flood is not associated with an earthquake.
Effects of a thermal gradient through the concrete section are not considered where the effects of the gradient reduce the effects caused by an abnormal loading condition.
Effects of a thermal gradient through the concrete section are not considered where the effects of the gradient reduce the effects caused by an abnormal loading condition.
Maximum values of time
Maximum values of time
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SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 31  Figure 3.8-13 illustrates the reinforced concrete idealization used in the model, and Figure 3.8-15 shows the shell
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 31  Figure 3.8-13 illustrates the reinforced concrete idealization used in the model, and Figure 3.8-15 shows the shell
-type force and moment resultants which are calculated for sections along the shell meridian, and subsequently used in the design of the reinforcement. Figure 3.8-16 illustrates the axisymmetric containment model with integral mat, wall and dome used in the first four analyses of Table 3.8-4. Representative node points and section numbers are shown where the latter range from 1 at the apex of the dome to 98 at the base of the wall. At any section, the model is composed of shell and solid elements representing the steel liner and reinforcement and concrete, respectively. Material properties are assigned to each layer according to the orientation of reinforcement and cracked or uncracked state of concrete. Force and moment resultants were computed for each load case by integrating stresses through the cross section. These cross-sectional forces and moments are combined for all applicable load combinations of Subsection 3.8.1.3. 2. Nonaxisymmetric Loads Nonaxisymmetric loads applied to the containment structure, other than localized loads, such as internal missiles, are the seismic and aircraft impact loads. The seismic input and seismic analysis are discussed in Subsections 3.7.1 and 3.7.2, respectively. The aircraft impact load and analysis are described in Appendix 2P. Wind and tornado loads are not considered because of the presence of the Enclosure Building. An elastic analysis of the aircraft impact loading was made using the Wilson II finite element code, "Dynamic Stress Analysis of Axisymmetric Structures Under Arbitrary Loading," by Ghosh and Wilson, as revised in September 1975. This code is further described in Subsection 3.8.1.4g and Appendix 3F. The loading was represented by Fourier series functions applied to the axisymmetric model of the containment.
-type force and moment resultants which are calculated for sections along the shell meridian, and subsequently used in the design of the reinforcement. Figure 3.8-16 illustrates the axisymmetric containment model with integral mat, wall and dome used in the first four analyses of Table 3.8-4. Representative node points and section numbers are shown where the latter range from 1 at the apex of the dome to 98 at the base of the wall. At any section, the model is composed of shell and solid elements representing the steel liner and reinforcement and concrete, respectively. Material properties are assigned to each layer according to the orientation of reinforcement and cracked or uncracked state of concrete. Force and moment resultants were computed for each load case by integrating stresses through the cross section. These cross-sectional forces and moments are combined for all applicable load combinations of Subsection 3.8.1.3. 2. Nonaxisymmetric Loads Nonaxisymmetric loads applied to the containment structure, other than localized loads, such as internal missiles, are the seismic and aircraft impact loads. The seismic input and seismic analysis are discussed in Subsections 3.7.1 and 3.7.2, respectively. The aircraft impact load and analysis are described in Appendix 2P. Wind and tornado loads are not considered because of the presence of the Enclosure Building. An elastic analysis of the aircraft impact loading was made using the Wilson II finite element code, "Dynamic Stress Analysis of Axisymmetric Structures Under Arbitrary Loading," by Ghosh and Wilson, as revised in September 1975. This code is further described in Subsection 3.8.1.4g and Appendix 3F. The loading was represented by Fourier series functions applied to the axisymmetric model of the containment.
The shear forces and overturning moments due to OBE and SSE seismic loads were obtained from the seismic analysis described in Subsection 3.7.2. These shear forces and moments were treated as effective static forces acting over the cross
The shear forces and overturning moments due to OBE and SSE seismic loads were obtained from the seismic analysis described in Subsection
 
====3.7.2. These====
shear forces and moments were treated as effective static forces acting over the cross
-section of the containment shell. Modeling assumptions for these two load cases are shown in Table 3.8-4.
-section of the containment shell. Modeling assumptions for these two load cases are shown in Table 3.8-4.
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 32  3. Design of Reinforcement The force and moment resultants for the individual load cases were combined for the load combinations of Subsection 3.8.1.3 and Article  
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 32  3. Design of Reinforcement The force and moment resultants for the individual load cases were combined for the load combinations of Subsection 3.8.1.3 and Article  
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: h. Tangential Shear The design of the containment for seismic loads incorporates diagonal shear reinforcing to resist tangential shear. The design for tangential shear is in accordance with the provisions of Article CC
: h. Tangential Shear The design of the containment for seismic loads incorporates diagonal shear reinforcing to resist tangential shear. The design for tangential shear is in accordance with the provisions of Article CC
-3000 of the ASME Code, with the exception of the limitation on allowable concrete shear stress shown as Note 1 of Table 3.8-2.
-3000 of the ASME Code, with the exception of the limitation on allowable concrete shear stress shown as Note 1 of Table 3.8-2.
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 36  The forces and moments due to seismic loading are discussed in Subsection 3.7.2. Three components of the seismic motion were assumed to exist simultaneously, and the resulting component forces and moments were combined by the squar e-root-of-the-sum-of-the squares (SRSS) method. The maximum tangential shear from the SRSS combination is assumed to act simultaneously at all points on the containment circumference at the given elevation. All forces and moments were combined per the specified load combinations. The LESCAL program, described in Subsection 3.8.1.4g, was used to calculate rebar stresses for all sections and elevations combining the stresses due to in
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 36  The forces and moments due to seismic loading are discussed in Subsection
 
====3.7.2. Three====
components of the seismic motion were assumed to exist simultaneously, and the resulting component forces and moments were combined by the squar e-root-of-the-sum-of-the squares (SRSS) method. The maximum tangential shear from the SRSS combination is assumed to act simultaneously at all points on the containment circumference at the given elevation. All forces and moments were combined per the specified load combinations. The LESCAL program, described in Subsection 3.8.1.4g, was used to calculate rebar stresses for all sections and elevations combining the stresses due to in
-plane forces and moments. Duchon's equations (Reference
-plane forces and moments. Duchon's equations (Reference
: 3) are incorporated into LESCAL for calculating rebar stresses (including inclined rebar) for the combined membrane forces. i. Variation in Physical Material Properties The effects of variations in material properties were considered in the design and analysis. Material properties which can strongly influence both analysis and design due to variability or uncertainty include:  (1) dynamic modulus of soils, (2) the modulus of elasticity of concrete and, (3) material strengths.
: 3) are incorporated into LESCAL for calculating rebar stresses (including inclined rebar) for the combined membrane forces. i. Variation in Physical Material Properties The effects of variations in material properties were considered in the design and analysis. Material properties which can strongly influence both analysis and design due to variability or uncertainty include:  (1) dynamic modulus of soils, (2) the modulus of elasticity of concrete and, (3) material strengths.
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: 2. The data will be obtained without embedding any devices in the cylinder or dome. A test report will be provided which will compare test results with predicted and allowable limits and evaluate any deviations.
: 2. The data will be obtained without embedding any devices in the cylinder or dome. A test report will be provided which will compare test results with predicted and allowable limits and evaluate any deviations.
Before and after the test, the containment will be visually examined to assure that no distress has developed on the concrete or liner.
Before and after the test, the containment will be visually examined to assure that no distress has developed on the concrete or liner.
3.8.2 Steel Containment The following sections contain the physical descriptions, applicable codes, standards and specifications, loads and load combinations, design and analysis procedures, allowable stresses, quality control and testing requirements for the major steel penetrations of the concrete containment structure that are pressure
 
-resisting but unbacked by concrete. The containment structure itself is constructed of reinforced concrete and, as such, is discussed in Subsection 3.8.1. 3.8.2.1 Description of the Containment Penetrations The containment penetrations described in this Subsection are the personnel air lock and equipment hatch, the fuel transfer tube assembly, the piping, electrical and instrumentation penetrations, and the ventilation penetrations. These components penetrate the containment shell to provide access, anchor piping, or furnish some other operational requirement. They also maintain leak tightness for the containment shell, and meet the maximum allowable leakage rate, as described in Subsection 3.8.1. Other functional characteristics are described below in the descriptions of individual components.
====3.8.2 Steel====
Containment The following sections contain the physical descriptions, applicable codes, standards and specifications, loads and load combinations, design and analysis procedures, allowable stresses, quality control and testing requirements for the major steel penetrations of the concrete containment structure that are pressure
-resisting but unbacked by concrete. The containment structure itself is constructed of reinforced concrete and, as such, is discussed in Subsection 3.8.1. 3.8.2.1 Description of the Containment Penetrations The containment penetrations described in this Subsection are the personnel air lock and equipment hatch, the fuel transfer tube assembly, the piping, electrical and instrumentation penetrations, and the ventilation penetrations. These components penetrate the containment shell to provide access, anchor piping, or furnish some other operational requirement. They also maintain leak tightness for the containment shell, and meet the maximum allowable leakage rate, as described in Subsection
 
====3.8.1. Other====
functional characteristics are described below in the descriptions of individual components.
All penetrations are anchored to sleeves (or to barrels) which are embedded in the concrete containment wall. This embedment is accomplished by means of an engineered anchorage system that is welded to the sleeve (or barrel) which is, in turn, welded to the locally thickened liner (see Subsection 3.8.1). a. Personal Air Lock (Breech Type)
All penetrations are anchored to sleeves (or to barrels) which are embedded in the concrete containment wall. This embedment is accomplished by means of an engineered anchorage system that is welded to the sleeve (or barrel) which is, in turn, welded to the locally thickened liner (see Subsection 3.8.1). a. Personal Air Lock (Breech Type)
The personnel air lock (Figure 3.8-20) consists of the air lock doors and the lock barrel. Its centerline is located at elevatio n 29'-6" and an azimuth of 315, as shown on Figure 1.2-4.
The personnel air lock (Figure 3.8-20) consists of the air lock doors and the lock barrel. Its centerline is located at elevatio n 29'-6" and an azimuth of 315, as shown on Figure 1.2-4.
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: 2. Both kinds of assemblies, however, were analyzed to meet the intent of Class MC criteria (NE) through the use of a simplified, but conservative, approach described in Subsection 3.8.2.4. d. Electrical Penetrations Electrical penetrations (Figure 3.8-24) consist of a stainless steel header plate (bulkhead) with an attached terminal box, electrical modules which are clamped to the header plate, and a carbon steel weld ring which is welded to the header plate and to the sleeve. The metallic pressure resisting parts, the sleeve, stainless steel header plate and carbon steel weld ring were designed as ASME III Safety Class MC components (NE); that portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). Double silicone and EPR O-rings provide a seal with a cavity for leakage monitoring between the header plate and the modules. The header plate is provided with a hole on the outside of the containment to allow for pressurization of the penetration assembly for leakage monitoring.
: 2. Both kinds of assemblies, however, were analyzed to meet the intent of Class MC criteria (NE) through the use of a simplified, but conservative, approach described in Subsection 3.8.2.4. d. Electrical Penetrations Electrical penetrations (Figure 3.8-24) consist of a stainless steel header plate (bulkhead) with an attached terminal box, electrical modules which are clamped to the header plate, and a carbon steel weld ring which is welded to the header plate and to the sleeve. The metallic pressure resisting parts, the sleeve, stainless steel header plate and carbon steel weld ring were designed as ASME III Safety Class MC components (NE); that portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). Double silicone and EPR O-rings provide a seal with a cavity for leakage monitoring between the header plate and the modules. The header plate is provided with a hole on the outside of the containment to allow for pressurization of the penetration assembly for leakage monitoring.


SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 55  e. Instrumentation Penetrations Instrumentation penetrations are of two types, electrical and fluid. The electrical type is similar in construction to the other electrical penetrations, and the discussion in Subsection 3.8.2.1d, as well as material appearing elsewhere in Subsection 3.8.2 which pertains to electrical penetrations, is applicable to these penetrations.
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 55  e. Instrumentation Penetrations Instrumentation penetrations are of two types, electrical and fluid. The electrical type is similar in construction to the other electrical penetrations, and the discussion in Subsection 3.8.2.1d, as well as material appearing elsewhere in Subsection
The second type of instrumentation penetrations, the fluid type, is similar in construction to the moderate energy piping penetrations. Consequently, the discussion in Subsection 3.8.2.1c on moderate energy piping penetration and other material in Subsection 3.8.2 relating to these penetrations are also applicable to this type of instrumentation penetration.
 
====3.8.2 which====
pertains to electrical penetrations, is applicable to these penetrations.
The second type of instrumentation penetrations, the fluid type, is similar in construction to the moderate energy piping penetrations. Consequently, the discussion in Subsection 3.8.2.1c on moderate energy piping penetration and other material in Subsection
 
====3.8.2 relating====
to these penetrations are also applicable to this type of instrumentation penetration.
: f. Fuel Transfer Tube Assembly The fuel transfer tube assembly (Figure 3.8-25) consists of the fuel transfer tube, the penetration sleeve, the fixed saddle on the reactor side, and the sliding saddle in the Fuel Storage Building. The fuel transfer tube and its flanges were designed as ASME III Class 2 components (NC). The quick closure hatch on the refueling canal side of the tube was designed and fabricated to ASME Section III Class MC component (NE) requirements. That portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). The remaining pieces of the assembly were designed as ASME III Component Supports (NF).
: f. Fuel Transfer Tube Assembly The fuel transfer tube assembly (Figure 3.8-25) consists of the fuel transfer tube, the penetration sleeve, the fixed saddle on the reactor side, and the sliding saddle in the Fuel Storage Building. The fuel transfer tube and its flanges were designed as ASME III Class 2 components (NC). The quick closure hatch on the refueling canal side of the tube was designed and fabricated to ASME Section III Class MC component (NE) requirements. That portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). The remaining pieces of the assembly were designed as ASME III Component Supports (NF).
: g. Ventilation Penetration Assemblies There are two types of ventilation penetrations, the containment air purge penetrations and the containment online purge penetrations. The containment air purge penetrations (Figure 3.8-26) each consist of a pipe sleeve (a rolled and welded pipe section, 36" O.D. by 1/2" wall thickness) which is flanged at each end with 36" weld neck flanges and, attached to these flanges, is a butterfly valve inside containment and a testable blind flange outside containment. Together with the pipe, the blind flanges form a part of the containment pressure boundary during plant Modes 1, 2, 3, and 4. During Modes 5 and 6, the blind flanges are replaced by spool pieces to configure the CAP system to perform its heating and ventilation functions. The valves are 36" diameter butterfly valves with fail
: g. Ventilation Penetration Assemblies There are two types of ventilation penetrations, the containment air purge penetrations and the containment online purge penetrations. The containment air purge penetrations (Figure 3.8-26) each consist of a pipe sleeve (a rolled and welded pipe section, 36" O.D. by 1/2" wall thickness) which is flanged at each end with 36" weld neck flanges and, attached to these flanges, is a butterfly valve inside containment and a testable blind flange outside containment. Together with the pipe, the blind flanges form a part of the containment pressure boundary during plant Modes 1, 2, 3, and 4. During Modes 5 and 6, the blind flanges are replaced by spool pieces to configure the CAP system to perform its heating and ventilation functions. The valves are 36" diameter butterfly valves with fail
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is equipped with a leak chase for pressure testing. Since the details, other than size, of these penetrations are essentially the same as those of the containment air purge penetrations, an additional figure is not included.
is equipped with a leak chase for pressure testing. Since the details, other than size, of these penetrations are essentially the same as those of the containment air purge penetrations, an additional figure is not included.
The portion of the penetration sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). That portion of the sleeve extending beyond the containment was designed to meet the requirements of ASME Section III, Division 1, Subsection NE, in addition to the fabrication and nondestructive examination requirements of Division
The portion of the penetration sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). That portion of the sleeve extending beyond the containment was designed to meet the requirements of ASME Section III, Division 1, Subsection NE, in addition to the fabrication and nondestructive examination requirements of Division
: 2. The valves and flanges were designed as ASME III Code Class 2 components (NC) and are further described in Subsection 3.9.3.
: 2. The valves and flanges were designed as ASME III Code Class 2 components (NC) and are further described in Subsection
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 57  3.8.2.2 Applicable Codes, Standards, and Specifications The design, materials, fabrication and inspection requirements for the major containment penetrations conform to, but are not necessarily limited to, the applicable sections of the following codes and specifications which are used to establish or implement design bases and methods, analytical techniques, material properties and quality control provisions.
 
====3.9.3. SEABROOK====
STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 57  3.8.2.2 Applicable Codes, Standards, and Specifications The design, materials, fabrication and inspection requirements for the major containment penetrations conform to, but are not necessarily limited to, the applicable sections of the following codes and specifications which are used to establish or implement design bases and methods, analytical techniques, material properties and quality control provisions.
Dates and revisions given for the listed codes are the earliest version that was used. Subsequent issues were incorporated into the design where practicable, or where the new issue directly affected the safety of the structure.
Dates and revisions given for the listed codes are the earliest version that was used. Subsequent issues were incorporated into the design where practicable, or where the new issue directly affected the safety of the structure.
Code or Specification Title ASME Boiler & Pressure Vessel Code Section II - Material Specification Section III, Division 1, Subsection NA, General Requirements Section III, Division 1, Subsection NB, Class 1 Components Section III, Division 1, Subsection NC, Class 2 Components Section III, Division 1, Subsection NE, Class MC Components Section III, Division 1, Subsection NF, Component Supports  Section III, Divisi on 2 Code for Concrete Reactor Vessels and Containments 1975 Edition including Winter 1976 Addendum for Containment Liner Section IX - Welding Qualifications
Code or Specification Title ASME Boiler & Pressure Vessel Code Section II - Material Specification Section III, Division 1, Subsection NA, General Requirements Section III, Division 1, Subsection NB, Class 1 Components Section III, Division 1, Subsection NC, Class 2 Components Section III, Division 1, Subsection NE, Class MC Components Section III, Division 1, Subsection NF, Component Supports  Section III, Divisi on 2 Code for Concrete Reactor Vessels and Containments 1975 Edition including Winter 1976 Addendum for Containment Liner Section IX - Welding Qualifications
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This tie reduces the torsional effect on the beam, which is designed to resist the remaining torsional effects. Lateral stability of the columns is provided by reinforced concrete beams and slabs connected to the secondary shield wall at Elevations 0' and 25'. The crane specification requires that the manufacturer's design insures that the crane will remain on the rail during either an OBE or an SSE. This is accomplished by means of kick
This tie reduces the torsional effect on the beam, which is designed to resist the remaining torsional effects. Lateral stability of the columns is provided by reinforced concrete beams and slabs connected to the secondary shield wall at Elevations 0' and 25'. The crane specification requires that the manufacturer's design insures that the crane will remain on the rail during either an OBE or an SSE. This is accomplished by means of kick
-back plates attached to the wheel carriage, which prevent separation between the wheel and the rail. See Figure 3.8-30. Seismic base shears are transmitted to the floor slab by means of a positive connection between the rail and the floor slab, which is part of the supporting structure. The crane rail anchor is designed to transfer all shear forces and overturning moments (resulting from the action of either an OBE or an SSE on the crane) to the concrete supporting system. Details of this connection include clamp plates, which restrain the rail and are welded to the base plate, and anchor bolts, which secure the base plate to the concrete beam. See Figure 3.8-30. m. Neutron Shielding The neutron shielding is a shell structure mounted on the slabs and walls above and around the reactor vessel. It consists of a plastic material encased in steel plate and provides radiation shielding.
-back plates attached to the wheel carriage, which prevent separation between the wheel and the rail. See Figure 3.8-30. Seismic base shears are transmitted to the floor slab by means of a positive connection between the rail and the floor slab, which is part of the supporting structure. The crane rail anchor is designed to transfer all shear forces and overturning moments (resulting from the action of either an OBE or an SSE on the crane) to the concrete supporting system. Details of this connection include clamp plates, which restrain the rail and are welded to the base plate, and anchor bolts, which secure the base plate to the concrete beam. See Figure 3.8-30. m. Neutron Shielding The neutron shielding is a shell structure mounted on the slabs and walls above and around the reactor vessel. It consists of a plastic material encased in steel plate and provides radiation shielding.
A more complete description of the neutron shielding and its effect on structures can be found in Section 6.2.
A more complete description of the neutron shielding and its effect on structures can be found in Section
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 95  3.8.3.2 Applicable Codes, Standards and Specifications The design, materials, fabrication and inspection of the internal structures conform to, but are not necessarily limited to, the applicable sections of the following codes and specifications which are used to establish or implement design bases and methods, analytical techniques, material properties and quality control provisions. Dates and revisions given for the listed codes are the earliest version that was used. Subsequent issues were incorporated into the design where practicable or where the new issue directly affected the safety of the structure.
 
===6.2. SEABROOK===
STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 95  3.8.3.2 Applicable Codes, Standards and Specifications The design, materials, fabrication and inspection of the internal structures conform to, but are not necessarily limited to, the applicable sections of the following codes and specifications which are used to establish or implement design bases and methods, analytical techniques, material properties and quality control provisions. Dates and revisions given for the listed codes are the earliest version that was used. Subsequent issues were incorporated into the design where practicable or where the new issue directly affected the safety of the structure.
Code or Specification Title  ACI 211.1-70  Recommended Practice for Selecting Proportions for Normal Weight Concrete ACI 214-65  Recommended Practice for Evaluation of Compression Test Results of Field Concrete ACI 301-72  Specification for Structural Concrete for Building (application sections)  ACI 302-69  Recommended Practice for Concrete Floor and Slab Construction ACI 304-73  Recommended Practice for Measuring, Mixing, Transporting and Placing Concrete ACI Committee Report 74-33  Recommended Practice for Hot Weather Concreting ACI Report 306R
Code or Specification Title  ACI 211.1-70  Recommended Practice for Selecting Proportions for Normal Weight Concrete ACI 214-65  Recommended Practice for Evaluation of Compression Test Results of Field Concrete ACI 301-72  Specification for Structural Concrete for Building (application sections)  ACI 302-69  Recommended Practice for Concrete Floor and Slab Construction ACI 304-73  Recommended Practice for Measuring, Mixing, Transporting and Placing Concrete ACI Committee Report 74-33  Recommended Practice for Hot Weather Concreting ACI Report 306R
-78  Recommended Practice for Cold Weather Concreting ACI 308-71  Recommended Practice for Curing Concrete ACI 309-72  Recommended Practice for Consolidation of Concret e  ACI Committee Placing Concrete by Pumping Method ACI Committee Report  
-78  Recommended Practice for Cold Weather Concreting ACI 308-71  Recommended Practice for Curing Concrete ACI 309-72  Recommended Practice for Consolidation of Concret e  ACI Committee Placing Concrete by Pumping Method ACI Committee Report  
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: a. Design and Analysis After preliminary design, the internal structures are analyzed to determine the maximum stress and displacements in reinforcing steel, concrete and structural steel for the loading criteria described in Subsection 3.8.3.3. Each structural system, either an individual component or a group of connected components, is analyzed separately using appropriate design assumptions and boundary conditions.
: a. Design and Analysis After preliminary design, the internal structures are analyzed to determine the maximum stress and displacements in reinforcing steel, concrete and structural steel for the loading criteria described in Subsection 3.8.3.3. Each structural system, either an individual component or a group of connected components, is analyzed separately using appropriate design assumptions and boundary conditions.


SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 115  The analyses include fixed base conditions, uncracked concrete, load superposition, modeling and selective interpretation of results. Material properties and boundary conditions are selected to be representative of the particular structure or component being analyzed. Either conventional methods and formulae of structural mechanics or the finite element direct stiffness method are used. All analyses consider elastic behavior only. The design is based on codes, specifications and documents listed in Subsection 3.8.3.2. The methods of design and analysis, including assumptions, boundary conditions, loads resisted and analysis techniques are given below. Seismic analysis is discussed in Subsection 3.7.2. Reinforcing in concrete components is proportioned in accordance with the responses obtained from the analyses using the strength design method of the ACI 318-71 code, as applicable, and the requirements of Subsections 3.8.3.3 and 3.8.3.5. Bond and anchorage requirements are in accordance with ACI 318-71. As a result, reinforcing patterns in walls and floors, in general, consist of orthogonal layers on both the inside and outside faces of the concrete.
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 115  The analyses include fixed base conditions, uncracked concrete, load superposition, modeling and selective interpretation of results. Material properties and boundary conditions are selected to be representative of the particular structure or component being analyzed. Either conventional methods and formulae of structural mechanics or the finite element direct stiffness method are used. All analyses consider elastic behavior only. The design is based on codes, specifications and documents listed in Subsection 3.8.3.2. The methods of design and analysis, including assumptions, boundary conditions, loads resisted and analysis techniques are given below. Seismic analysis is discussed in Subsection
 
====3.7.2. Reinforcing====
in concrete components is proportioned in accordance with the responses obtained from the analyses using the strength design method of the ACI 318-71 code, as applicable, and the requirements of Subsections 3.8.3.3 and 3.8.3.5. Bond and anchorage requirements are in accordance with ACI 318-71. As a result, reinforcing patterns in walls and floors, in general, consist of orthogonal layers on both the inside and outside faces of the concrete.
The computer program CONCOL (described in Subsection 3.8.3.4) was used throughout the internals for ultimate strength design of concrete. Other computer programs used in the design and analysis of specific internal structures are described below.
The computer program CONCOL (described in Subsection 3.8.3.4) was used throughout the internals for ultimate strength design of concrete. Other computer programs used in the design and analysis of specific internal structures are described below.
Structural steel is designed in accordance with the AISC Specification, using elastic methods, except as noted below for the reactor support system.
Structural steel is designed in accordance with the AISC Specification, using elastic methods, except as noted below for the reactor support system.
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: d. Steel and Concrete Coating System Materials used for coating the internal structures are the same as those described in Subsection 3.8.1.6. These materials meet the requirements of ANSI Standard N101.2. 3.8.3.7 Testing and In
: d. Steel and Concrete Coating System Materials used for coating the internal structures are the same as those described in Subsection 3.8.1.6. These materials meet the requirements of ANSI Standard N101.2. 3.8.3.7 Testing and In
-Service Surveillance Requirements Quality control testing as discussed in Subsection 3.8.3.6 will be employed. No additional testing or in-service surveillance is required.
-Service Surveillance Requirements Quality control testing as discussed in Subsection 3.8.3.6 will be employed. No additional testing or in-service surveillance is required.
3.8.4 Other Seismic Category I Structures This section contains physical descriptions, codes, loads and load combinations, design and analysis procedures, allowable stresses, quality control, and testing requirements as they relate to seismic Category I structures exclusive of the containment structure and its internals. Nonseismic Category I structures which received special design considerations to prevent collapse onto adjacent Category I structures are also included in this section.
 
====3.8.4 Other====
Seismic Category I Structures This section contains physical descriptions, codes, loads and load combinations, design and analysis procedures, allowable stresses, quality control, and testing requirements as they relate to seismic Category I structures exclusive of the containment structure and its internals. Nonseismic Category I structures which received special design considerations to prevent collapse onto adjacent Category I structures are also included in this section.
3.8.4.1 Description of the Structures Locations of all of the major plant structures are shown on the plant arrangement plan, Figure 1.2-1; overall dimensions of the seismic Category I structures are given in Table 3.8-15.
3.8.4.1 Description of the Structures Locations of all of the major plant structures are shown on the plant arrangement plan, Figure 1.2-1; overall dimensions of the seismic Category I structures are given in Table 3.8-15.
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 126  All seismic Category I structures are separated from adjacent structures above the point of fixity by means of isolation joints, except for a few structures described below in which two portions of the same structure are used for two different functions. The resulting multi
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 126  All seismic Category I structures are separated from adjacent structures above the point of fixity by means of isolation joints, except for a few structures described below in which two portions of the same structure are used for two different functions. The resulting multi
Line 2,022: Line 2,073:
6 feet and a 4'
6 feet and a 4'
-10" thick concrete floor; for leak tightness the inside surface is lined with stainless steel plates 3/16" thick on the walls and 1/4" thick on the floor. The building superstructure is constructed of concrete walls and roof.
-10" thick concrete floor; for leak tightness the inside surface is lined with stainless steel plates 3/16" thick on the walls and 1/4" thick on the floor. The building superstructure is constructed of concrete walls and roof.
New and spent fuel storage and handling are described in Section 9.1. Refer to Figur e 1.2-15, Figure 1.2-16, Figure 1.2-17, Figure 1.2-18, Figure 1.2-19, Figure 1.2-20 and Figure 1.2-21 for plan and sectional elevations of the building.
New and spent fuel storage and handling are described in Section
 
===9.1. Refer===
to Figur e 1.2-15, Figure 1.2-16, Figure 1.2-17, Figure 1.2-18, Figure 1.2-19, Figure 1.2-20 and Figure 1.2-21 for plan and sectional elevations of the building.


SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 129  i. Main Steam and Feedwater Pipe Chase (East) Including East Penetration Area The main steam and feedwater pipe chase (east) is a reinforced concrete structure which houses and protects the main steam and feedwater piping. The overall dimensions are 127.58 feet long, 22.25 feet wide for most of the length, and 61.5 feet high. It is 41.25 feet wide at the widest point.
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 129  i. Main Steam and Feedwater Pipe Chase (East) Including East Penetration Area The main steam and feedwater pipe chase (east) is a reinforced concrete structure which houses and protects the main steam and feedwater piping. The overall dimensions are 127.58 feet long, 22.25 feet wide for most of the length, and 61.5 feet high. It is 41.25 feet wide at the widest point.
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See Figure 3.8-34 and Figure 3.8-35 for the duct banks and manholes respectively.
See Figure 3.8-34 and Figure 3.8-35 for the duct banks and manholes respectively.
: o. Service Water Cooling Tower Including Switchgear Rooms The Service Water Cooling Tower is a rectangular building approximately 300 feet x 54 feet in plan, extending 28 feet below grade, and projecting 57.5 feet above ground. It is located on the south side of the plant and is founded on rock. The cooling tower serves as an ultimate heat sink in the unlikely event that the cooling water tunnels are rendered inoperative. The cooling tower houses pumps, fans, water distribution system and nozzles.
: o. Service Water Cooling Tower Including Switchgear Rooms The Service Water Cooling Tower is a rectangular building approximately 300 feet x 54 feet in plan, extending 28 feet below grade, and projecting 57.5 feet above ground. It is located on the south side of the plant and is founded on rock. The cooling tower serves as an ultimate heat sink in the unlikely event that the cooling water tunnels are rendered inoperative. The cooling tower houses pumps, fans, water distribution system and nozzles.
The switchgear rooms are approximately 54 feet x 26 feet in plan and extend from Elevation 22.0 feet to Elevation 46.0 feet. They are located on the east and west ends of the cooling tower and are an integral part of the structure. The switchgear rooms house the switchgear, substation, and motor control center for the cooling tower. See Figure 1.2-56. Operational characteristics of the cooling tower are described in Subsection 9.2.5.
The switchgear rooms are approximately 54 feet x 26 feet in plan and extend from Elevation 22.0 feet to Elevation 46.0 feet. They are located on the east and west ends of the cooling tower and are an integral part of the structure. The switchgear rooms house the switchgear, substation, and motor control center for the cooling tower. See Figure 1.2-56. Operational characteristics of the cooling tower are described in Subsection
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 132  p. Service Water Pumphouse The Service Water Pumphouse (SWPH) is located on the plant site east of the Containment Building and is founded on rock. It is attached to and shares a foundation with the Circulating Water Pumphouse (CWPH), a non
 
====9.2.5. SEABROOK====
STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 132  p. Service Water Pumphouse The Service Water Pumphouse (SWPH) is located on the plant site east of the Containment Building and is founded on rock. It is attached to and shares a foundation with the Circulating Water Pumphouse (CWPH), a non
-Category I structure which is designed so that its loss or failure will not impair the Service Water Pumphouse or System.
-Category I structure which is designed so that its loss or failure will not impair the Service Water Pumphouse or System.
The structure under the SWPH is a reinforced concrete basin and is approximately  
The structure under the SWPH is a reinforced concrete basin and is approximately  
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-related equipment and components located in the electrical tunnels.
-related equipment and components located in the electrical tunnels.


SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 143  In the Waste Processing Building only those areas housing radioactive gaseous waste equipment were designed for tornado loads; the other areas do not contain safety-related equipment. Procedures by which the structures were checked for missile loads, tornado-generated as well as internally generated, are described in Section 3.5. Determination of pressures on structures due to tornado is described in Subsection 3.3.2. Pressure loadings from wind are described in Subsection 3.3.1. The seismic analysis of seismic Category I structures is described in Subsection 3.7.2. All cranes in these structures are furnished with hold
SEABROOK STATION UFSAR  DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 143  In the Waste Processing Building only those areas housing radioactive gaseous waste equipment were designed for tornado loads; the other areas do not contain safety-related equipment. Procedures by which the structures were checked for missile loads, tornado-generated as well as internally generated, are described in Section 3.5. Determination of pressures on structures due to tornado is described in Subsection
 
====3.3.2. Pressure====
loadings from wind are described in Subsection 3.3.1. The seismic analysis of seismic Category I structures is described in Subsection 3.7.2. All cranes in these structures are furnished with hold
-down devices to ensure that they are not dislodged by earthquake forces. Monorails, by nature of their support mechanisms, cannot be dislodged by earthquake forces.
-down devices to ensure that they are not dislodged by earthquake forces. Monorails, by nature of their support mechanisms, cannot be dislodged by earthquake forces.
Using methods outlined in TID
Using methods outlined in TID
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Code Class Seismic Category Building (11) Supplier Notes (14)  Reactor Coolant System Reactor Vessel        Vessel and Closure Head 1 ASME III (7) 1 I CS W  Full Length CRDM Housing 1 ASME III 1 I CS W  CRDM Head Adapter Plugs 1 ASME III 1 I CS W  5.3 Incore Instrumentation Guide Tubes 1 ASME III 1 I CS W  Reactor Coolant Pump        Pump Casing 1 ASME III 1 I CS W  Thermal Barrier 1 ASME III 1 I CS W  Thermal Barrier Heat  
Code Class Seismic Category Building (11) Supplier Notes (14)  Reactor Coolant System Reactor Vessel        Vessel and Closure Head 1 ASME III (7) 1 I CS W  Full Length CRDM Housing 1 ASME III 1 I CS W  CRDM Head Adapter Plugs 1 ASME III 1 I CS W  5.3 Incore Instrumentation Guide Tubes 1 ASME III 1 I CS W  Reactor Coolant Pump        Pump Casing 1 ASME III 1 I CS W  Thermal Barrier 1 ASME III 1 I CS W  Thermal Barrier Heat  


Exchanger 1 ASME III 1 I CS W No. 1 Seal Housing 1 ASME III 1 I CS W  No. 2 Seal Housing 2 ASME III 1 I CS W See Note 4. Pressure Retaining        Bolting 1 ASME III 1 I CS W  5.4.1 Pump Flywheel 3 N/A N/A I CS W Steam Generator        Tube (Reactor Coolant) Side 1 ASME III 1 I CS W 5.4.2 Shell (Steam) Side 2 ASME III 1 I CS W See Note 4.
Exchanger 1 ASME III 1 I CS W No. 1 Seal Housing 1 ASME III 1 I CS W  No. 2 Seal Housing 2 ASME III 1 I CS W See Note 4. Pressure Retaining        Bolting 1 ASME III 1 I CS W  5.4.1 Pump Flywheel 3 N/A N/A I CS W Steam Generator        Tube (Reactor Coolant) Side 1 ASME III 1 I CS W
 
====5.4.2 Shell====
(Steam) Side 2 ASME III 1 I CS W See Note 4.
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Sheet: 13 2 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
Sheet: 13 2 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
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Valves(pressurized) 2 ASME III 2 I CS AE  Containment Recirculating Filter System        Fans 3 AMCA - I CS AE  Filter Unit NNS ANSI N509/
Valves(pressurized) 2 ASME III 2 I CS AE  Containment Recirculating Filter System        Fans 3 AMCA - I CS AE  Filter Unit NNS ANSI N509/
N510 - - CS AE  Ductwork and Dampers 3 SMACNA, ANSI N509/N510 - I CS AE  Containment Purge System        All pressurized components upstream of emergency exhaust filters (flow meter, throttle valve, piping) 2 ASME III 2 I CS/CE AE  Containment Atmosphere Sample System        Supply and Exhaust Piping 2 ASME III 2 I CS/CE AE 6.2.5 Supply and Exhaust Valves 2 ASME III 2 I CS/CE AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
N510 - - CS AE  Ductwork and Dampers 3 SMACNA, ANSI N509/N510 - I CS AE  Containment Purge System        All pressurized components upstream of emergency exhaust filters (flow meter, throttle valve, piping) 2 ASME III 2 I CS/CE AE  Containment Atmosphere Sample System        Supply and Exhaust Piping 2 ASME III 2 I CS/CE AE
 
====6.2.5 Supply====
and Exhaust Valves 2 ASME III 2 I CS/CE AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Sheet: 13 7 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
Sheet: 13 7 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  


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Code Class Seismic Category Building (11) Supplier Notes (14) Emergency Core Cooling  
Code Class Seismic Category Building (11) Supplier Notes (14) Emergency Core Cooling  


System        Accumulator 2 ASME III 2 I PB W Safety Injection Pump 2 ASME III 2 I PB W See Notes 1a, 1c & 2. Piping        Reactor Coolant Pressure Boundary 1 ASME III 1 I CS/PB W See Note 3. Drain, vent and test lines beyond the first pressure boundary isolation valve. NNS ANSI B31.1 - - PB AE 6.3 Valves 2 ASME III and ANSI B16.5 2 I PB/CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
System        Accumulator 2 ASME III 2 I PB W Safety Injection Pump 2 ASME III 2 I PB W See Notes 1a, 1c & 2. Piping        Reactor Coolant Pressure Boundary 1 ASME III 1 I CS/PB W See Note 3. Drain, vent and test lines beyond the first pressure boundary isolation valve. NNS ANSI B31.1 - - PB AE
 
===6.3 Valves===
2 ASME III and ANSI B16.5 2 I PB/CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Sheet: 13 8 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
Sheet: 13 8 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  


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N510 - I FB AE  Ductwork & Dampers 3 ANSI N509/
N510 - I FB AE  Ductwork & Dampers 3 ANSI N509/
N510 - I FB AE  Control Room Emergency Filtration Subsystem Fans 3 AMCA - I CD AE/PSNH  Filters (HEPA & Charcoal) 3 ANSI N509/
N510 - I FB AE  Control Room Emergency Filtration Subsystem Fans 3 AMCA - I CD AE/PSNH  Filters (HEPA & Charcoal) 3 ANSI N509/
N510 - I CD AE/PSNH 6.5.1 Ductwork and Dampers 3 ANSI N509/
N510 - I CD AE/PSNH
 
====6.5.1 Ductwork====
and Dampers 3 ANSI N509/
N510 - I CD AE/PSNH S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
N510 - I CD AE/PSNH S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Sheet: 13 9 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
Sheet: 13 9 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
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Loop 3 ASME III 3 I FB AE See Note 3. Spent Fuel Pool Skimmer  
Loop 3 ASME III 3 I FB AE See Note 3. Spent Fuel Pool Skimmer  


Loop NNS ANSI B31.1 - - FB/PB AE 9.1.3 Refueling Canal Cleanup  
Loop NNS ANSI B31.1 - - FB/PB AE
 
====9.1.3 Refueling====
Canal Cleanup  


Loop NNS ANSI B31.1 - - CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Loop NNS ANSI B31.1 - - CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
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Code Class Seismic Category Building (11) Supplier Notes (14) Station Service Water  
Code Class Seismic Category Building (11) Supplier Notes (14) Station Service Water  


System        Service Water Pumps 3 ASME III 3 I CW AE Piping and Valves 3 ASME III 3 I SW/PB/YD AE 9.2.1 Strainers 3 ASME III 3 I PB AE  Primary Component  
System        Service Water Pumps 3 ASME III 3 I CW AE Piping and Valves 3 ASME III 3 I SW/PB/YD AE
 
====9.2.1 Strainers====
3 ASME III 3 I PB AE  Primary Component  


Cooling Water System (PCCW)        PCCW Pumps 3 ASME III 3 I PB AE  PCCW Heat Exchanger 3 ASME III 3 I PB AE  PCCW Head Tank 3 ASME III 3 I PB AE  Piping and Valves        Furnish and support cooling water supply to safeguards components 2/3 ASME III 2/3 I CS/CF/FS AE See Note 3. Containment Penetration 2 ASME III 2 I CE AE Other NNS ANSI B31.1 - - PB/CS WB/FB/YD AE  Thermal Barrier        Pumps 3 ASME III 3 I CS AE  HX 2/3 ASME III 2/3 I CS AE  9.2.2 Head Tank 3 ASME III 3 I CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Cooling Water System (PCCW)        PCCW Pumps 3 ASME III 3 I PB AE  PCCW Heat Exchanger 3 ASME III 3 I PB AE  PCCW Head Tank 3 ASME III 3 I PB AE  Piping and Valves        Furnish and support cooling water supply to safeguards components 2/3 ASME III 2/3 I CS/CF/FS AE See Note 3. Containment Penetration 2 ASME III 2 I CE AE Other NNS ANSI B31.1 - - PB/CS WB/FB/YD AE  Thermal Barrier        Pumps 3 ASME III 3 I CS AE  HX 2/3 ASME III 2/3 I CS AE  9.2.2 Head Tank 3 ASME III 3 I CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
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Code Class Seismic Category Building (11) Supplier Notes (14) Ultimate Heat Sink Atlantic  
Code Class Seismic Category Building (11) Supplier Notes (14) Ultimate Heat Sink Atlantic  


Ocean        Intake & Discharge Tunnels NNS - - - - AE  Piping 3 ASME III and 3 I YD AE NNS ANSI B31.1 - - YD/TB AE  Expansion Joints 3 ASME III and - - YD AE NNS ANSI B31.1 3 1 PA/SW AE  Cooling Tower        Cooling Tower Fans 3 MFRS. STDS. - I CT AE  Cooling Tower Pumps 3 ASME III 3 I CT AE  9.2.5 Piping and Valves 3 ASME III 3 I CT AE Condensate Storage Facilities        Condensate Storage Tank 3 ASME III 3 I YD AE 9.2.6 Piping and Valves NNS ANSI B31.1 - - YD/TB AE Reactor Makeup Water  
Ocean        Intake & Discharge Tunnels NNS - - - - AE  Piping 3 ASME III and 3 I YD AE NNS ANSI B31.1 - - YD/TB AE  Expansion Joints 3 ASME III and - - YD AE NNS ANSI B31.1 3 1 PA/SW AE  Cooling Tower        Cooling Tower Fans 3 MFRS. STDS. - I CT AE  Cooling Tower Pumps 3 ASME III 3 I CT AE  9.2.5 Piping and Valves 3 ASME III 3 I CT AE Condensate Storage Facilities        Condensate Storage Tank 3 ASME III 3 I YD AE
 
====9.2.6 Piping====
and Valves NNS ANSI B31.1 - - YD/TB AE Reactor Makeup Water  


System        Reactor Makeup Water NNS API-650, - - YD AE Storage Tank  ASME VIII Reactor Makeup Water Pump NNS ANSI B73.1 - - PB AE  Piping and Valves        Containment Penetration 2 ASME III 2 I CS AE  9.2.7 Other NNS ANSI B31.1 - - PB/WB/CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
System        Reactor Makeup Water NNS API-650, - - YD AE Storage Tank  ASME VIII Reactor Makeup Water Pump NNS ANSI B73.1 - - PB AE  Piping and Valves        Containment Penetration 2 ASME III 2 I CS AE  9.2.7 Other NNS ANSI B31.1 - - PB/WB/CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Line 3,508: Line 3,588:


Ventilation System        Ventilation Fans  AMCA - - PB AE  PCCW and BI Pump Area Wall  AMCA - I PB AE  Fans and Dampers Normal Cleanup Exhaust Fans  AMCA - - PB AE  Normal Cleanup Filters  ANSI N509/
Ventilation System        Ventilation Fans  AMCA - - PB AE  PCCW and BI Pump Area Wall  AMCA - I PB AE  Fans and Dampers Normal Cleanup Exhaust Fans  AMCA - - PB AE  Normal Cleanup Filters  ANSI N509/
N510 - - PB AE  Ductwork and Dampers Isolation Dampers and Ductwork Passing through CE 2 ANSI N509/N510 - I PB/CE AE 9.4.3 Other NNS SMACNA - I PB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
N510 - - PB AE  Ductwork and Dampers Isolation Dampers and Ductwork Passing through CE 2 ANSI N509/N510 - I PB/CE AE
 
====9.4.3 Other====
NNS SMACNA - I PB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Sheet: 13 25 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
Sheet: 13 25 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  


Line 3,516: Line 3,599:
Ventilation System        Fans - AMCA - - WB AE  Ductwork - SMACNA - - WB AE  Dampers - AMCA - - WB AE  9.4.4 Filters - ANSI N509/N510 - - WB AE Containment Ventilation  
Ventilation System        Fans - AMCA - - WB AE  Ductwork - SMACNA - - WB AE  Dampers - AMCA - - WB AE  9.4.4 Filters - ANSI N509/N510 - - WB AE Containment Ventilation  


System        Recirculation Fan/Coolers - MFRS. STDS. - I CS AE See Note 9.
System        Recirculation Fan/Coolers - MFRS. STDS. - I CS AE See Note 9.  
9.4.5 Ductwork - SMACNA - I CS AE See Note 10.
 
====9.4.5 Ductwork====
- SMACNA - I CS AE See Note 10.
Containment Enclosure  
Containment Enclosure  


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Code Class Seismic Category Building (11) Supplier Notes (14) Diesel Generator Building  
Code Class Seismic Category Building (11) Supplier Notes (14) Diesel Generator Building  


Ventilation System        Fans 3 AMCA - I CD AE  Ductwork 3 (Mod) SMACNA - I CD AE 9.4.8 Dampers 3 AMCA - I CD AE Cable Spreading Room  
Ventilation System        Fans 3 AMCA - I CD AE  Ductwork 3 (Mod) SMACNA - I CD AE
 
====9.4.8 Dampers====
3 AMCA - I CD AE Cable Spreading Room  


Ventilation System        Fans - AMCA - - CD AE  Ductwork - SMACNA - I CD AE  9.4.9 Dampers - AMCA - I CD AE  4 kV Switchgear Area and Battery Rooms Ventilation  
Ventilation System        Fans - AMCA - - CD AE  Ductwork - SMACNA - I CD AE  9.4.9 Dampers - AMCA - I CD AE  4 kV Switchgear Area and Battery Rooms Ventilation  
Line 3,550: Line 3,638:


Codes/Stds.
Codes/Stds.
Code Class Seismic Category Building (11) Supplier Notes (14) Fire Protection System        Water Supply Tank - NFPA STDS. - - YD AE  Fire Pump - NFPA STDS. - - YD AE  Jockey Pump - NFPA STDS. - - YD AE 9.5.1 Piping and Valves - NFPA STDS. - - YD AE Diesel Generator Fuel Oil Storage and Transfer  
Code Class Seismic Category Building (11) Supplier Notes (14) Fire Protection System        Water Supply Tank - NFPA STDS. - - YD AE  Fire Pump - NFPA STDS. - - YD AE  Jockey Pump - NFPA STDS. - - YD AE
 
====9.5.1 Piping====
and Valves - NFPA STDS. - - YD AE Diesel Generator Fuel Oil Storage and Transfer  


System        Fuel Oil Day Tank 3 ASME III 3 I CD AE  Fuel Oil Storage Tank 3 ASME III 3 I CD AE  Fuel Oil Transfer Pump 3 ASME III 3 I CD AE Associated Piping and Valves for Above Tanks and Pump 3 ASME III 3 I CD AE 9.5.4 All Remaining On-Engine Equipment and Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.
System        Fuel Oil Day Tank 3 ASME III 3 I CD AE  Fuel Oil Storage Tank 3 ASME III 3 I CD AE  Fuel Oil Transfer Pump 3 ASME III 3 I CD AE Associated Piping and Valves for Above Tanks and Pump 3 ASME III 3 I CD AE 9.5.4 All Remaining On-Engine Equipment and Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.
Line 3,564: Line 3,655:
System        Air Receivers 3 ASME III 3 I CD AE  Air Compressor 3 MFRS. STDS. - I CD AE  Piping and Valves from Air  
System        Air Receivers 3 ASME III 3 I CD AE  Air Compressor 3 MFRS. STDS. - I CD AE  Piping and Valves from Air  


Receiver to DG Skid 3 ASME III 3 I CD AE  All Remaining Essential Equipment Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.
Receiver to DG Skid 3 ASME III 3 I CD AE  All Remaining Essential Equipment Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.  
9.5.6 Backup Control Air Compressor 3 MFRS. STDS. - I CD S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
 
====9.5.6 Backup====
Control Air Compressor 3 MFRS. STDS. - I CD S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Sheet: 13 30 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
Sheet: 13 30 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  


Line 3,576: Line 3,669:
Combustion Air Intake and  
Combustion Air Intake and  


Exhaust System        Piping (Off-Engine) 3 ANSI B31.1 - I CD AE  Air Intake Filter 3 MFRS. STDS. - I CD AE  Exhaust Silencer 3 MFRS. STDS. - I CD AE  Exhaust Expansion Joints 3 MFRS. STDS. - I CD AE  Piping (On-Engine) 3 MFRS. STDS. - I CD AE 9.5.8 Steam & Power Conversion System S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Exhaust System        Piping (Off-Engine) 3 ANSI B31.1 - I CD AE  Air Intake Filter 3 MFRS. STDS. - I CD AE  Exhaust Silencer 3 MFRS. STDS. - I CD AE  Exhaust Expansion Joints 3 MFRS. STDS. - I CD AE  Piping (On-Engine) 3 MFRS. STDS. - I CD AE
 
====9.5.8 Steam====
& Power Conversion System S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:
Sheet: 13 31 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  
Sheet: 13 31 of 42  UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.  


Line 3,696: Line 3,792:


S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.5-1 Revision:
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.5-1 Revision:
Sheet: 8 1 of 1  TABLE 3.5-1 S TRUCTURES , SHIELDS A ND B ARRIERS D ESIGNED T O R ESIST I NTERNAL A ND EXTERNAL M ISSILES  Internal Missiles External Missiles Containment shell 1 Containment structure primary shield wall 2  Secondary shield wall 2  Pressurizer shielding 2  Reactor missile shield 2  Operating floor 2  Primary Auxiliary Building 3 1 Refueling water storage tank See Section 6.2 Condensate storage tank 1 Control and Diesel Generator Building 3 1 Shield between diesel generators 3  Fuel Storage Building 3 1 Electrical penetration tunnel 1 Service Water Pumphouse, intake and discharge transition structures, flume and backwash ducts 3 1 Emergency Feedwater Pump Building 3 1 Diesel fuel tanks Piping tunnels 3 1 Main steam and feedwater lines outside containment to 2nd isolation valve 1 Category I electrical manholes 1  1 - Tornado generated missiles 2 - Missiles postulated within the containment 3 - Temperature sensing element missile  
Sheet: 8 1 of 1  TABLE 3.5-1 S TRUCTURES , SHIELDS A ND B ARRIERS D ESIGNED T O R ESIST I NTERNAL A ND EXTERNAL M ISSILES  Internal Missiles External Missiles Containment shell 1 Containment structure primary shield wall 2  Secondary shield wall 2  Pressurizer shielding 2  Reactor missile shield 2  Operating floor 2  Primary Auxiliary Building 3 1 Refueling water storage tank See Section  
 
===6.2 Condensate===
storage tank 1 Control and Diesel Generator Building 3 1 Shield between diesel generators 3  Fuel Storage Building 3 1 Electrical penetration tunnel 1 Service Water Pumphouse, intake and discharge transition structures, flume and backwash ducts 3 1 Emergency Feedwater Pump Building 3 1 Diesel fuel tanks Piping tunnels 3 1 Main steam and feedwater lines outside containment to 2nd isolation valve 1 Category I electrical manholes 1  1 - Tornado generated missiles 2 - Missiles postulated within the containment 3 - Temperature sensing element missile  


S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-2 Revision:
S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-2 Revision:
Line 4,343: Line 4,442:
228  7,670 Base Hold-down bolts, Tension, psi 3,450 19,100  Shear, psi 2,430  9,900 Deflections Pump shaft, in.
228  7,670 Base Hold-down bolts, Tension, psi 3,450 19,100  Shear, psi 2,430  9,900 Deflections Pump shaft, in.
0.0038 0.022  Coupling parallel misalignment, in.
0.0038 0.022  Coupling parallel misalignment, in.
0.0385 0.102  Coupling angular misalignment, degrees 0.264 1.5 Motor rotor, in.
0.0385 0.102  Coupling angular misalignment, degrees 0.264  
 
===1.5 Motor===
rotor, in.
0.0011 0.043 S EABROOK S TATION UFSA R D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-17 Revision:
0.0011 0.043 S EABROOK S TATION UFSA R D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-17 Revision:
Sheet: 9 1 of 1  TABLE 3.9(B)-17 STRESS AND DEFLECTION
Sheet: 9 1 of 1  TABLE 3.9(B)-17 STRESS AND DEFLECTION
Line 4,666: Line 4,768:
Sheet:    8 1 of 1  TABLE 3.9(N)
Sheet:    8 1 of 1  TABLE 3.9(N)
-9 STRESS CRITERIA FOR SAFETY-RELATED ASME CODE CLASS 2 A ND CLASS 3 VALVES  Design/Service Level Stress Limits (Notes 1
-9 STRESS CRITERIA FOR SAFETY-RELATED ASME CODE CLASS 2 A ND CLASS 3 VALVES  Design/Service Level Stress Limits (Notes 1
-4) Pmax (Note 5) Design & Service Valve bodies shall conform 1.0 Level A    to ASME Section III Service Level B m  1.1 S  1.1        m Lb  1.65 S Service Level C m  1.5 S  1.2        m Lb  1.80 S Service Level D m  2.0 S  1.5        m Lb  2.4 S Notes: 1. Valve nozzle (piping load) stress analysis is not required when both the following conditions are satisfied: (1) the section modulus and area of energy plane, normal to the flow, through the region defined as the valve body crotch are at least 110% of those for the piping connected (or join
-4) Pmax (Note 5) Design & Service Valve bodies shall conform
 
===1.0 Level===
A    to ASME Section III Service Level B m  1.1 S  1.1        m Lb  1.65 S Service Level C m  1.5 S  1.2        m Lb  1.80 S Service Level D m  2.0 S  1.5        m Lb  2.4 S Notes: 1. Valve nozzle (piping load) stress analysis is not required when both the following conditions are satisfied: (1) the section modulus and area of energy plane, normal to the flow, through the region defined as the valve body crotch are at least 110% of those for the piping connected (or join
-ed) to the valve body inlet and outlet nozzles; and, (2) code allowable stress, S, for valve body material is equal to or greater than the code allowable stress, S, for connected piping material. If the valve body material allowable stress is less than that of the connected  
-ed) to the valve body inlet and outlet nozzles; and, (2) code allowable stress, S, for valve body material is equal to or greater than the code allowable stress, S, for connected piping material. If the valve body material allowable stress is less than that of the connected  


Latest revision as of 07:11, 11 October 2018

Seabrook Station Updated Final Safety Analysis Report, Revision 16, Chapter 3, Design of Structures, Components, Equipment and Systems
ML14317A509
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 10/21/2014
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To:
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S EABROOK S TATION U PDATED F INAL S AFETY A NALYSIS R EPORT C HAPTER 3 DESIGN OF STRUCTURES

, COMPONENTS

, EQUIPMENT AND SYSTEMS

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 1 3.1 CONFORMANCE TO NRC GE NERAL DESIGN CRITERIA This section discusses the extent to which the design criteria for plant structures, systems and components important to safety meet the NRC Ge neral Design Criteria for Nuclear Power Plants, specified in Appendix A to 10 CFR Part 50. For each criterion, a summary is provided which discusses how the principal design features meet the criterion and identifies any exceptions that are taken. In the discussion of each criterion, th e chapters or sections of this Updated FSAR where more detailed information is presented are referenced to demonstrate compliance with the criterion.

3.1.1 Overall

Requirements 3.1.1.1 Criterion 1 - Quality Standards and Records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keep ing with th e required safety function. A quality assurance program sh all be established and implemented to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropr iate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licen see throughout the life of the unit.

RESPONSE The structures, systems and components of the facility are classified according to their importance in the prevention and mitigation of accidents, using the classification system developed by the American Nuclear Society. Each safety-related component is given a safety class designation. The codes, standards and quality controls applicable to each type of safety-related component and its safety class designation are identified in Chapter 3. Where applicable, design, fabrication, er ection and testing are in accordance with the codes required in 10 CFR 50.55a.

Recognized codes and standards, when used, are identified and evaluated to assure their applicability, adequacy and sufficiency in keeping with the requi red safety function.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 2 The quality assurance programs established by Yankee Atomic Electric Company, United Engineers and Constructors Inc., and Westinghouse conform with the requirements of 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, and are discussed in Chapter 17. Chapter 14 describes initial tests and operation to assure performance on installed equipment commensurate with the importance of the safety function.

Component safety classifications are shown on P&IDs presented w ith their appropriate sections. In accordance with the quality assurance programs, complete sets of records of the design, fabrication, construction and testing of safety-related components are maintained by the nuclear power licensee throughout the life of the plant. 3.1.1.2 Criterion 2 - Design Bases for Pr otection Against Natural Phenomena Structures, systems, and components important to safety shall be desi gned to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflec t: (1) appropriate consideration of the most severe of the natural phenomena that have been historically reported fo r the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.

RESPONSE The structures, systems and components important to safety are designed to either withstand the effects of natural phenome na without loss of capability to perform their safety functions, or to fail in the safest condition. Those structures, systems and components vital to the safe shutdown capability of the reactor are designed to withstand the maximum probable natural phenomenon expected at the site, determined from recorded da ta for the site vicinity, with appropriate margin to account for uncertainties in historical data. Appropriate combinations of normal, accident and natural phenomena structural loadings are considered in the design of the structures, systems and components important to safety. The nature and magnitudes of the natural phenomena considered in the de sign of this plant are discussed in Sections 2.3, 2.4, and 2.5. Sections 3.2 through 3.11 discuss the design of the plant in relationship to natural events. Seismic and safety classifications, as well as other pertinent standards and information, are given in the s ections discussing individual structures and components.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 3 3.1.1.3 Criterion 3 - Fire Protection Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capab ility shall be provided and designed to minimize the effects of fires on structures, systems, and co mponents important to safety. Fire-fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

RESPONSE The plant is designed to minimize the probability and effect of fires and explosions. Noncombustible and fire-resistant materials are used in the containment and control room, as well as in components of safety features systems and wherever practic al throughout the plant. Equipment and facilities for fire protection, including detection, alarm, and extinguishment are provided to protect both plant equipment and personnel from fire and explosion. Fire-suppression systems are designed to assure that their ruptur e or inadvertent operation will not significantly impair systems important to safety.

The Fire Protection System is described in Chapter 9. 3.1.1.4 Criterion 4 - Environmental and Missile Design Bases Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 4 RESPONSE Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing and postulated accidents, including loss-of-coolant accidents. Criteria are presented in Chapter 3; environmental conditions ar e presented in Chapter 3 and 6. These structures, systems, and components are pr otected, as appropriate, against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids that may result from equipment failures and from events and c onditions outside the nuclear power unit. Details of the design, environmental testing, and construction of these systems, structures, and components are included in Chapters 3, 5, 6, 7 and 9. Evaluation of the performance of safety features is contai ned in Chapter 15. 3.1.1.5 Criterion 5 - Sharing of Structures, Systems, and Components Structures, systems, and components important to safety shall not be shared among nuclear power plants unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

RESPONSE Since Seabrook Station is a single unit plant, there is no sharing of structures, systems and/or components as described in GDC-5.

Further discussion is presen ted in Chapters 6, 8, and 9. 3.1.2 Protection by Multiple Fission Product Barriers 3.1.2.1 Criterion 10 - Reactor Design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated oper ational occurrences.

RESPONSE The reactor core and associated coolant, control, and protection systems are designed with adequate margins to:

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 5 Assure that fuel damage

  • is not expected during normal core operation and operational transients (Condition I)
    • or any transient conditions arising from occurrences of moderate frequency (Condition II)
    • . It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the Plant Cleanup System, and are consistent with plant design bases. Ensure return of the reactor to a safe state following a Condition III
    • event with only a small fraction of fuel rods damaged, although sufficient fuel damage might occur to preclude resumption of operation.

Assure that the core is intact with acceptable heat transfer geometry following transients arising from occurrences of limiting faults (Condition IV)

    • . Chapter 4 discusses the design bases and design evaluation of core components. Details of the control and protection systems instrumentation design and logic are discussed in Chapter 7. This information supports the accident analyses of Chapter 15, which show that the acceptable fuel design limits are not exceeded for Condition I and II occurrences. 3.1.2.2 Criterion 11 - Reactor Inherent Protection The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

RESPONSE Prompt compensatory reactivity feedback effects are assured when the reactor is critical by the negative fuel temperature effect (Doppler effect) and by the non-positive operational limit on moderator temperature coefficient of reactivity. A non-negative moderator temperature coefficient is allowed by Technical Specifications for all power levels, provided that compliance with the ATWS Rule and its basis are maintained, as described in the Bases for Technical

Specification 3/4.1.1.3. The Seabrook core design philosophy meets this requirement by ensuring that a non-positive MTC exists for operating conditions above 20% power. The negative Doppler coefficient of reactivity is assured by the inherent design using low-enrichment fuel; the non-positive moderator temperature coefficient of reactiv ity is assured by administratively controlling the dissolved abso rber concentration or by burnable poison.

  • Fuel damage as used here is defined as a penetration of the fission product barrier (i.e., the fuel rod clad).

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 6 These reactivity coefficients are discussed in Section 4.3. 3.1.2.3 Criterion 12 - Suppression of Reactor Power Oscillations The reactor core and associated coolant, control, and protection system shall be designed to assure that power oscillations which can result in conditions exceeding sp ecified acceptable fuel design limits are not possible or can be re liably and readily de tected and suppressed.

RESPONSE Power oscillations of the fundamental mode are inherently eliminated by the negative Doppler and non-positive moderator temperature coefficient of reactivity. Oscillations, due to xenon spatial effects, in the radial, diametral, and azimuthal overtone modes are heavily damped due to the inherent design and due to the negative Doppler and non-positive moderator temperature coefficients of reactivity.

Oscillations, due to xenon spatial effects, in the axial first overtone mode may occur. Assurance that fuel design limits are not exceeded by xenon axial oscillations is provided by reactor trip functions using the measured axial power imbalance as an input. Oscillations, due to xenon spatial effects, in axial modes higher than the first overtone, are heavily damped due to the inherent design a nd due to the negative D oppler coefficient of reactivity.

Xenon stability control is described in Section 4.3. 3.1.2.4 Criterion 13 - Instrumentation and Control Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for antici pated operational oc currences and accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 7 RESPONSE Instrumentation and controls are provided to monitor and c ontrol neutron flux, control rod position, temperatures, pressures, flows, and levels as necessary to assure that adequate plant safety can be maintained. Instrumentation is provided in the Reactor Coolant System, Steam and Power Conversion System, the Containment, Engineered Safety Features Systems, Radiological Waste Systems and other auxiliaries. Parameters that must be provided for operator use under normal operating and accident conditi ons are indicated in the control room in proximity with the controls for maintaining the indicated parameter in the proper range. Appropriate controls maintain these variables and systems within prescribed operating ranges. The quantity and types of process instrumentation provided ensures safe and orderly operation of all systems over the full design range of the plant. These systems are described in Chapter 6, 7, 8, 9, 11 and 12. 3.1.2.5 Criterion 14 - Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be desi gned, fabricated, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. RESPONSE The Reactor Coolant System boundary is designed to accommodate the system pressures and temperatures attained under all expected modes of plant operation, including all anticipated transients, and to maintain the stresses within applicable stress limits. See Section 3.9 for details. Reactor coolant pressure boundary materials, sel ection and fabrication techniques ensure a low probability of gross rupture or significant leakage.

In addition to the loads imposed on the system under normal operating conditions, consideration is also given to abnormal loading conditions, such as seismic and pipe rupture as discussed in Sections 3.6 and 3.7. The system is protected from overpressure by means of pressure relieving devices as required by applicab le codes (see Subsection 5.2.2). The Reactor Coolant System boundary has provisions for inspection, testing and surveillance of critical areas to assess the structural and leak-tight integrity. See Section 5.2 for details. For the reactor vessel, a material surveillance program conforming to applicable codes is provided. See Section 5.3 for details.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 8 3.1.2.6 Criterion 15 - Reactor Coolant System Design The Reactor Coolant System and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

RESPONSE The design pressure and temperature for each component in the reactor coolant and associated auxiliary control, and protection systems are selected to be above the maximum coolant pressure and temperature under all normal and an ticipated transient load conditions. Additionally, reactor coolant pressure boundary components achieve a large margin of safety by

using proven ASME materials and design codes, proven fabrication tec hniques, nondestructive shop testing, and integrated hydrostatic testing of assembled components. Chapter 5 discusses the reactor coolant system design. 3.1.2.7 Criterion 16 - Containment Design Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions importa nt to safety are not ex ceeded for as long as postulated accident conditions require.

RESPONSE A reinforced concrete outer containment enclosure encloses the steel-line d reinforced concrete primary containment structure which, in turn, encl oses the entire Reactor Coolant System. The primary containment structure is designed to sustain, without loss of integrity, all effects of the most limiting faults up to and including the double-ended rupture of the largest pipe in the Reactor Coolant System. The Emergency Core Cooling System and the Containment Spray System serve to cool the reactor core and the containment and return the containment to near atmospheric pressure following a LOCA. Leakage from the primary containment is limite d by design during the post LOCA pressurized period. Continued primary containment integr ity is assured by pe riodic leak testing. The secondary containment provides the ability to process any leakage from the primary containment, which results from th e post LOCA differential pressure.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 9 The containment and safety-related systems ar e designed to assure the required functional capability of containing any uncontrolled release of radioactivity, and no design conditions important to safety are exceeded for as long as postulated accident conditions require.

Refer to Chapters 1, 3, 6, 15 and 16. 3.1.2.8 Criterion 17 - Electrical Power Systems An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

The onsite electric power supplies, including the batteries, and the Onsite Electric Distribution System, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.

Electric power from the transmission network to the Onsite Electric Distribution System shall be supplied by two physically independent circuits (not necessari ly on separate rights of way) designed and located to minimize to the extent practical the lik elihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alte rnating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few second s following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained. Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmissi on network, or the loss of power from the onsite electric power supplies.

RESPONSE An onsite electric power system is provided a nd is designed with adequate independence, capacity, redundancy, and testability to assure the functioning of safety-related systems. Independence is provided by physical separation of components and cables to minimize vulnerability of redundant engineered safety features systems to single credible accidents.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 10 Offsite power from the transmission network to the switching substation is supplied by three physically independent transmissi on lines. Power from the switching substation is provided for the safety features system of each unit through both the unit auxiliary transformers and reserve auxiliary transformers. These two transforme r banks are independently connected to the switching substation and to each safety features load group to provide tw o circuits to power the two redundant load groups within a few seconds following a loss-of-coolant accident. A failure of a single component will not prevent safety-related systems from performing their function.

Chapter 8 provides further details.

The onsite AC source of electrical power, described in Section 8.3, consists of two diesel generators, one connected to each of the redundant emergency buses of the unit. One diesel generator is capable of supplyi ng sufficient power for the operation of the minimum safety features required for the unit during a postulated loss-of-coolant accident.

However, during a postulated loss-of-coolant accident, each diesel generator starts automatically and, if offsite power is not available, it connects to its associated emergenc y bus. The safety features equipment is then sequentially started. The emergency buses and their associated diesel generators are so arranged that a failure of a single component will not prevent the power supply systems from performing their function.

Chapter 8 provides further details.

Four DC batteries are provided in physically separated rooms, and two of these (one in each train) are adequate to supply the DC control power required for the safety features. For each of the four protection channels, an independent 120V AC vital power source is provided. Two additional vital 120V AC power sour ces are provided for other safety-related loads. Failure of a single component in this system will not impair control of the minimum safety features required to maintain the unit in a safe condition. Chapter 8 provides further details. 3.1.2.9 Criterion 18 - Inspection and Testing of Electrical Power Systems Electrical power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation seque nce that brings the systems into operation, including operation of applicable portions of the Protection System, and the transfer of power among the nuclear power unit, the Offsite Power System, and the Onsite Power System.

RESPONSE Class 1E systems are designed to permit appr opriate periodic inspec tion and testing of:

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 11 a. Important areas and features such as wiring, insulation, connections, and switchboards during equipment shutdown. b. The operability and functional performance of power system components, such as diesel generators, relays, buses, their DC system and circuit breakers and control circuits. c. The operability of the electric power systems as a whole during plant shutdown. d. The full operational sequence that brings the power systems into operation under conditions as close to design as practic al, including operation of the Protection System and the transfer of power among the nuclear power unit, the preferred (offsite) power supply, and the standby (onsite) power supply systems. Complete provisions for testing of Class 1E electric power systems and the standby power supplies (diesel generators) are described in Chapter 8. 3.1.2.10 Criterion 19 - Control Room A control room shall be provided from which acti ons can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protections shall be provided to permit access and occupancy of the control room under accident condi tions without personnel receiving radiation exposures in excess of 5 rem whole body, or its eq uivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room sha ll be provided (1) with a design capability for prompt hot shutdown of the unit in a safe condition during hot shutdown, and (2) with a potentia l capability for subsequent cold shutdown of the re actor through the use of suitable procedures.

RESPONSE Safe occupancy capability of the control room during normal and accident conditions is provided in the design. The control room is in a seismic Category I structure. Adequate shielding is provided to maintain radiation levels in the control room below 5 rem whole body, or its equivalent to any part of the body, for the dur ation of a postulated design basis accident. The control room is equipped with the primary a nd secondary control panels that contain those instruments and controls necessary to operate the plant safely under normal conditions and maintain it in a safe condition under accidents, including loss-of-coolant accidents.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 12 In the unlikely event that access to the control room is restricted, local control stations and manual operation of critical components can be used to effect hot shutdown from outside the control room for an extended period. By use of appropriate emergency procedures, the unit can also be brought to cold shutdown from outside the control room.

Refer to Chapters 1, 6, 7, 9 and 13 for specific details. 3.1.3 Protection and Reactivity Control Systems 3.1.3.1 Criterion 20 - Protection System Functions The Protection System shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety. RESPONSE A fully automatic protection system with appropria te redundant channels is provided to cope with transients where insufficient time is available for manual correctiv e action. The design basis for all protection systems is in accordance with the intent of IEEE Standard 279-1971 and IEEE Standard 379-1972. The Reactor Protection System automatically initiates a reactor trip when any variable monitored by the system or combination of monitored variables exceed the normal operating range. Setpoints are designed to provide an envelope of safe operating conditions with adequate margin for uncertainties to ensure that fuel design limits are not exceeded. Reactor trip is initiated by removing power to the rod drive mechanisms of all the full length Rod Cluster Control Assemblies. This causes the r ods to insert by gravit y rapidly reducing the reactor power output. The response and adequacy of the protection system has been verified by analysis of anticipated transients.

The Engineered Safety Features (ESF) Actuation System automatically initiates emergency core cooling, and other safeguards functions, by sens ing accident conditions using redundant analog channels measuring diverse variables. Manual actuation of safeguards may be performed where ample time is available for operator action. Th e ESF Actuation System automatically trips the reactor on manual or automatic sa fety injection signal generation. The Protection System is discussed in Chapter 7.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 13 3.1.3.2 Criterion 21 - Protection Syst em Reliability and Testability The Protection System shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed. Redundancy and independence designed into the Protection System shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from servic e of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the Protection System can be otherwise demonstrated. The Protection System shall be designed to permit periodic testing of its functioning when the r eactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

RESPONSE The Protection System is designed for high functional reliability and in-service testability. The Protection System meets the single failure requirements of IEEE Standard 279-1971 during normal operation. The rationale th at justifies the exception to th e single failure criterion during channel bypass for one-out-of-two systems is equally applicable to the two-out-of-three and two-out-of-four systems addressed in WCAP-10271. See Subsections 7.2.2 and 7.3.2 for a detailed discussion of comp liance with this criterion. 3.1.3.3 Criterion 22 - Protection System Independence The Protection System shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and pr inciples of operation, shall be us ed to the extent practical to prevent loss of the protection function.

RESPONSE Protection system components are designed and arranged so that the environment accompanying any emergency situation in which the components ar e required to function doe s not result in loss of the safety function. Various means are used to accomplish this. Functional diversity has been designed into the system. The extent of this f unctional diversity has been evaluated for a wide variety of postulated accidents.

Diverse protection functions will automatically terminate an accident before intolerable consequences can occur. See Sections 7.2 and 7.3 for details.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 14 High quality components, conservative design a nd applicable quality control, inspection, calibration, and tests are used to guard against common-mode failure. Qualification testing is performed on the various safety systems to demonstrate functional operation at normal and post-accident conditions of temperature, humidity, pressure, and radiation for specified periods, as required. The test results indicate no loss of the protection function. Refer to Sections 3.10 and 3.11 for further details. 3.1.3.4 Criterion 23 - Protecti on System Failure Modes The Protection System shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

RESPONSE The Protection System is designed with due consideration of the most probable failure modes of the components under various perturbations of the environment and energy sources. Each reactor trip channel is designed on the de-energize-to-trip principle so that loss of power, disconnection, open channel faults, and the majority of internal channel shor t-circuit faults cause the channel to go into its tripped mode. The Protection System is discussed in Sections 7.2 and 7.3. 3.1.3.5 Criterion 24 - Separation of Protection and Control Systems The Protection System shall be separated from control systems to the extent that failure of a single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliabil ity, redundancy, and independence requirements of the Protection System. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

RESPONSE The Protection System is separate and distinct from the Control Systems. Control Systems may be dependent on the Protection System in that control signals are derived from protection system measurements, where applicable. These signals are transferre d to the Control System by isolation amplifiers which are classified as protection components. The adequacy of system isolation has been verified by te sting under conditions of postulate d credible faults. The failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the Control and Protection System leaves intact a system which satisfies the requirements of the Protection System.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 15 The Protection System is discussed in Chapter 7. 3.1.3.6 Criterion 25 - Protection System Requirements for Reactivity Control Malfunctions The Protection System shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

RESPONSE The Protection System is designed to limit reactivity transients so that design limits are not exceeded. Reactor shutdown by full length rod insertion is completely independent of the normal control function since the trip breakers interrupt power to the rod mechanisms regardless of existing control signals. Thus in the postulated accidental withdrawal, (assumed to be initiated by a control malfunction) flux, temperature, pressu re, level and flow signals would independently be generated. Any of these signals (trip demands) would operate the breakers to trip the reactor.

Analyses of the effects of possible malfunctions are discussed in Chapter 15. These analyses show that for postulated dilution during refueling, startup, or manual or automatic operation at power, the operator has ample time to determine the cause of dilution, terminate the source of dilution and initiate reboration before the shutdown margin is lost. The analyses show that acceptable fuel damage limits are not exceeded even in the event of a single malfunction of either system. The Protection System is discussed in Chapter 7. 3.1.3.7 Criterion 26 - Reactivity Contro l System Red undancy and Capability Two independent reactivity control systems of diffe rent design principals shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipa ted operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcriti cal under cold conditions.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 16 RESPONSE Two Reactivity Control Systems are provided. These are Rod Cluster Control Assemblies (RCCAs) and chemical shim (boric acid). The R CCAs are inserted into the core by the force of gravity. During operation the shutdown rod banks are fully withdrawn. The full length Control Rod System automatically maintains a programmed average reactor temperature compensating for reactivity effects associated with scheduled and transient load changes.

The shutdown rod banks along with the full length control banks are de signed to shutdown the re actor with adequate margin under conditions of normal operation and anticipated operationa l occurrences thereby ensuring that specified fuel design limits are not exceeded. The most restrictive period in core life is assumed in all analyses and the most reactive rod cluster is assumed to be in the fully

withdrawn position. Sufficient boron can be provided to maintain the reactor in the cold shutdown state independent of the position of the control rods and can compensate for xenon burnout transients. Details of the construction of the RCCAs are presented in Chapter 4 and the operation is discussed in Chapter 7. The means of controlli ng the boric acid concentr ation is described in Chapter 9. Performance analyses under accid ent conditions are included in Chapter 15. 3.1.3.8 Criterion 27 - Combined Reactivity Control Systems Capability The Reactivity Control Systems shall be designed to have a combined capability, in conjunction with poison addition by the Emergency Core Cooling System, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

RESPONSE The facility is provided with means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. These means are discussed in detail in Chapters 4 and 9. Combin ed use of the Rod Cluster Control System and the Chemical Shim Control System permits the necessary shutdown margin to be maintained during long-term xenon decay and plant cooldown. The single highest worth control cluster is assumed to be stuck full-out upon trip for this determination.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 17 3.1.3.9 Criterion 28 - Reactivity Limits The Reactivity Control Systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support st ructures or other reactor vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

RESPONSE The maximum reactivity worth of control rods and the maximum rates of reactivity insertion employing control rods are limited to values that prevent rupture of the reactor coolant pressure boundary or disruptions of the core or vessel internals to a degree that could impair the effectiveness of emergency core cooling.

The maximum positive reactivity insertion rates for the withdrawal of Rod Cluster Control Assemblies (RCCAs) and the dilution of the boric acid in the Reactor Coolant System are limited by the physical design characteristics of the RCCAs and the Chemical and Volume Control System. Technical Specifications on shutdown margin and on RCCA insertion limits and bank overlaps as functions of power provide additional assurance that the consequences of the postulated accidents are no more se vere than those presented in the analyses of Chapter 15.

Reactivity insertion rates, dilution and withdrawal limits are also discussed in Section 4.3. The capability of the Chemical and Volume Control System to avoid an inadvertent excessive rate of boron dilution is discussed in Chapter 15.

Assurance of core cooling capability following Condition IV accidents, such as rod ejection, steam line break, etc., is given by keeping the reactor coolant pressure bounda ry stresses within faulted condition limits, as specified by applicable ASME codes. Structural deformations are checked also, and limited to values that do no t jeopardize the operation of necessary safety features. 3.1.3.10 Criterion 29 - Protection Against Anticipated Operational Occurrences The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 18 RESPONSE The protection and reactivity control systems are designed to assure extremely high probability of performing their required safety functions in the event of anticipated operational occurrence. Likely failure modes of system components are designed to be safe modes. Equipment used in these systems is designed, constructed, operated and maintained with a high level of reliability. Loss of power to the Protection System results in a reactor trip. Details of system design are covered in Chapter 7. Also refer to the discussions of GDC-20 through 25. 3.1.4 Fluid Systems 3.1.4.1 Criterion 30 - Quality of Re actor Coolant Pressure Boundary Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards pr actical. Means shall be provided for detecting and, to the extent practical, identifying the location of the sour ce of reactor coolant

leakage. RESPONSE Reactor coolant pressure boundary components are designed, fabricated, inspected and tested in conformance with ASME Boiler and Pressure Vesse l Code Sec. III, Rules for Construction of Nuclear Power Plant Components. All components are classified according to ANSI N 18.2A-1975 and are accorded the quality measures appropriate to the classification. The design bases and evaluations of reactor coolant pressure boundary components are discussed in Chapter 5.

Leakage is detected by an increase in the amount of makeup water required to maintain a normal level in the pressurizer. The reactor vessel closure joint is provided with a temperature-monitored leak-off between double gaskets. Leakage into the reactor containment is collected in the containment building sump where it is monitored. Leakage is also detected by measuring the airbor ne activity. Monitoring th e inventory of rector coolant in the system at the pressurizer, volume control tank and coolant drain collection tanks make available an accurate assessment of integrated leakage.

The Reactor Coolant Pressure Boundary Leakage Detection System is discussed in Subsection 5.2.5.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 19 3.1.4.2 Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary The reactor coolant pressure bounda ry shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testi ng, and postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) th e probability of rapidly propagating fracture is minimized. The design shall re flect consideration of service temperatures and other conditions of the boundary material under operating, main tenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual steady-state and transient stresses, and (4) size of flaws. RESPONSE Close control is maintained over material se lection and fabrication for the Reactor Coolant System to assure that the boundary behaves in a non-brittle manner. The Reactor Coolant System materials which are exposed to the cool ant are corrosion resistan t stainless steel or Inconel. The reference temperature (RT NDT) of the reactor vessel structural steel is established by Charpy V-notch and drop weight tests in accordance with 10 CFR 50, Appendix G.

As part of the reactor vessel specification, ce rtain requirements which are not specified by the applicable ASME Codes ar e performed, as follows: 1. Ultrasonic Testing - In addition to code requirements, the performance of a 100 percent volumetric ultrasonic test of reactor vessel plate for shear wave and a post-hydro test ultrasonic map of all welds in the pressure vessel are required. Cladding bond ultrasonic inspection to more restrictive requirements than those specified in the code are also required to preclude interpretation problems during in-service inspection. 2. Radiation Surveillance Program - In the surveillance programs, the evaluation of the radiation damage is based on pre-ir radiation and post-irra diation testing of Charpy V-notch and tensile specimens. These programs are directed toward evaluation of the effect of radiation on the fracture toughness of reactor vessel steels based on the reference transition temperature approach and the fracture mechanics approach, and are in accordance with ASTM E185-79, "Recommended Practi ce for Surveillance Tests for Nuclear Reactor Vessels," and the requirements of 10 CFR 50, Appendix H. 3. Reactor vessel core region material chemistry (copper, phosphorous and vanadium) is controlled to reduce sensitivity to embrittlement due to irradiation over the life of the plant.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 20 The fabrication and quality control techniques used in the fabrication of the Reactor Coolant System are equivalent to those used for the reactor vessel. The inspections of reactor vessel, pressurizer, piping, pumps, and steam generator are governed by ASME Code requirements. See Chapter 5 for details. Allowable pressure-temperature relationships for plant heatup and cooldown rates are calculated using methods derived from the ASME Code,Section III, Appendix G, "Protection Against Non-Ductile Failure." The approach specifies that allowed stress intensity factors for all vessel operating conditions shall not exceed the reference stress intensity factor (KIR) for the metal temperature at any time. Operating specifications include conservative margins for predicted changes in the material reference temperatures (RTNDT) due to irradiation. 3.1.4.3 Criterion 32 - Inspection of Reactor Coolant Pressure Boundary Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leak-tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

RESPONSE The design of the reactor coolant pressure boundary provides the capability for accessibility during service life to the entire internal surfaces of the reactor vessel, certain external zones of the vessel including the nozzle to reactor coolant piping welds and the top and bottom heads, and external surfaces of the reactor coolant piping except for the area of pipe within the primary shielding concrete. The inspection capability complements the leakage detection systems in assessing the pressure boundary component's integrity. The reactor coolant pressure boundary will be periodically inspected under the provisions of ASME Boiler and Pressure Vessel Code,Section XI. Details of the in-service inspection programs are pres ented in Subsection 5.2.4.

Monitoring of changes in the fracture toughness pr operties of the reactor vessel core region plates, forging, weldments and associated heat treated zones are performed in accordance with 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements." Samples of reactor vessel plate materials are retained and catal ogued in case future engineering development shows the need for further testing. The material properties surveillance program includes not only the conventional tensile and impact tests, but also fracture mechanics specimens. The observed shifts in RT NDT of the core region materials with irradiation will be used to confirm the allowable limits calculated for all operational transients. See S ection 5.3 for further details.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 21 3.1.4.4 Criterion 33 - Reactor Coolant Makeup A system to supply reactor coolant makeup for pr otection against small br eaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation.

RESPONSE The Chemical and Volume Control System provides a means of reactor coolant makeup and adjustment of the boric acid conc entration. Makeup is added automatically if the level in the volume control tank falls below the preset level. The high-pressure centrifugal charging pumps provided are capable of supplying the required makeup and reactor coolant seal injection flow when power is available from either onsite or offsite electric power systems. These pumps also serve as high head injection pumps. Functiona l reliability is assure d by provision of standby components assuring a safe response to probable modes of failure. Details of system design are included in Sections 6.3 and 9.3, with details of the Electric Power System included in Chapter 8. 3.1.4.5 Criterion 34 - Residual Heat Removal A system to remove residual heat shall be provided. The system safety function shall be to

transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, a nd suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

RESPONSE The Residual Heat Removal System, in conj unction with the Steam and Power Conversion System, is designed to transfer the fission production decay heat and other residual heat from the reactor core within acceptable limits. Initiation of the residu al heat removal system occurs at approximately 350 F and 425 psig.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 22 Suitable redundancy at temperatures below 350F is accomplished with two RHR drop lines, the two residual heat removal pumps (located in separate compartments with means available for draining and monitoring of leakage), the two heat exchangers and the associated piping, cabling and electric power sources. The Residual Heat Removal System is able to operate on either onsite or offsite electrical power systems. Suitable redundancy at temperatures above approximately 350F is provided by the four steam generators and attendant piping system. Details of the system design are in Subsection 5.4.7 and Chapter 10, with details of the Electric Power System included in Chapter 8. 3.1.4.6 Criterion 35 - Emergency Core Cooling A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. Suitable redundancy in components and features, a nd suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

RESPONSE An emergency core cooling system is provided to cope with any loss-of-c oolant accident in the plant design basis. Abundant cooling water is available in an em ergency to transfer heat from the core at a rate sufficient to maintain the core in a coolable geometry, and to assure that clad metal-water reaction is limited to less than one percent. Adequate design provisions are made to assure performance of the required safety functions even with a single failure. Details of the capability of the system are included in Section 6.3 with details of the Electric Power System included in Chapter 8. An evaluation of the adequacy of the system functions is included in Chapter 15. Performance evaluations are conducted in accordance with 10 CFR Part 50.46 and Appendix K.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 23 3.1.4.7 Criterion 36 - Inspection of Emergency Core Cooling System The Emergency Core Cooling System shall be designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping to assure integrity and capability of the system.

RESPONSE Design provisions facilitate access to the critica l parts of the injection piping and valves for visual inspection and for nondestructive inspection where such techniques are desirable and appropriate. The design is in accordance with ASME,Section XI requirements. The components outside the containment are accessible for leak-tightness inspection during operation of the reactor.

Details of the inspection program for the Emergency Core Cooling System are included in Sections 6.3, 6.6 and Chapter 16. 3.1.4.8 Criterion 37 - Testing of Emergency Core Cooling System The Emergency Core Cooling System shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the Protection System, the transfer between normal and emergency power sources, and the operation of the Associated Cooling Water System.

RESPONSE Each active component of the Emergency Core Cooling System (ECCS) may be individually actuated on the normal power source or transferred to the emergency power source at any time during appropriate plan t periodic tests. Tests may be performed during shutdown to demonstrate proper automatic operation of the ECCS, and assure the structural and leak-tight integrity of the system components. The details of the ECCS testing program are included in Section 6.3 and Chapter 16. The emergency power system testing program is described in Chapter 8.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 24 3.1.4.9 Criterion 38 - Containment Heat Removal A system to remove heat from the reactor containment shall be provided. The system function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels. Suitable redundancy in components and features, a nd suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

RESPONSE A containment spray system is provided to remove heat from the reactor containment following an accident. Its function is to maintain the pressure and temperature of the containment below the design values at all times. The system consists of two independent identical subsystems supplied from separate power buses. No single failure causes loss of more than half of the installed 200 percent cooling capacity. The system is described in Subsection 6.2.2.

Electrical facilities are described in Chapter 8. 3.1.4.10 Criterion 39 - Inspection of Co ntainment Heat Removal System The Containment Heat Removal System shall be designed to permit appropriate periodic inspection of important components, such as th e torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system.

RESPONSE The Containment Heat Removal System is accessible for inspection, as described below: a. The containment spray headers and nozzles are accessible for inspection when the reactor is shut down or at zero power. b. The containment spray heat exchanger, spray pumps and associated valves are located in an enclosure adjacent to the reactor containment. These components are readily available for inspection at all times.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 25 c. The containment recirculation sump strainer external surface is accessible for inspection. If necessary, the pump inlet plenum can be accessed using remote equipment when the reactor is shut down.

d. The refueling water storage and spray additive tank is readily accessible for inspection at all times.

For details, see Subsection 6.2.2. 3.1.4.11 Criterion 40 - Testing of Co ntainment Heat Removal System The Containment Heat Removal System shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and, under conditions as close to the design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the Protection System, the transfer between normal and emergency power sources, and the operation of the Associated Cooling Water System.

RESPONSE System piping, valves, pumps, heat exchanger, and other components of the Containment Heat Removal Systems are arranged so that each compone nt can be tested periodically for operability. The delivery capability of the Containment Spray System is tested periodically, to the extent practical, up to the last isolation valve before the spray nozzles. The delivery capability of the spray nozzles is tested periodically by blowing low-pressure air through the nozzles and verifying flow. The Containment Spray System is tested for operational sequence as close to the design as practical. For details, see Chapter 6 for the testing of the Containment Heat Removal System; Chapter 8 for the testing of the Emergency Power System. 3.1.4.12 Criterion 41 - Containment Atmosphere Cleanup Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be pr ovided as necessary to reduce, consistent with the functioning other associated systems, the concentration and quantity of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 26 Each system shall have suitable redundance in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure.

RESPONSE In the event of a loss-of-coolant accident, the Containment Spray System sprays a basic sodium hydroxide-borate solution into the containment atmosphere to remove fission product iodine. The spray system consists of two independent subsystems, each supplied from separate buses. Either subsystem alone can provide the iodine removal capacity for which credit is taken in Chapter 15.

The Post-Accident Hydrogen Control System is designed with redundant components so that a single failure does not prevent operation of the system. This system is described in Chapter 6. A hydrogen mixing system is provided, comprised of redundant fans and ductwork to prevent the concentration of hydrogen. This system is described in Chapter 9.

The details of the Electric Power System are included in Chapter 8.

No single failure causes the subsystems to fail to function. The secondary containment is maintained at a negative pressure, and acts as a collection system for containment leakage, which is collected and filtered prior to discharge to the environment. 3.1.4.13 Criterion 42 - Inspection of Cont ainment Atmosphere Cleanup Systems The Containment Atmosphere Cleanup Systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems.

RESPONSE The Containment Atmosphere Cleanup Systems are designed and located so that they can be inspected periodically, as required. The spray headers and nozzles can be air-tested, as described in the response to Criterion 39.

The containment combustible gas control system components can be inspected and periodically tested.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 27 See Chapter 6 for details. 3.1.4.14 Criterion 43 - Testing of Cont ainment Atmosphere Cleanup Systems The Containment Atmosphere Cleanup Systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, incl uding operation of appli cable portions of the Protection System, the transfer between normal and emergency power sources, and the operation of associated systems.

RESPONSE The Containment Atmosphere Cleanup System can be tested as follows: a. Operation of the spray pumps can be tested by recirculation to the refueling water storage tank through a test line. b. System valves can be operated through their full travel. c. Leak-tightness of the system is checked during testing. See Chapter 6 for details.

Power transfer is described in Chapter 8. 3.1.4.15 Criterion 44 - Cooling Water A system to transfer heat from structures, systems, and compone nts important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions. Suitable redundancy in components and features, a nd suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operations (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 28 RESPONSE The Cooling Water Systems for safety-related systems and components consist of the Primary Component Cooling Water System, the Service Water System and the ultimate heat sink. Each system has two trains that are capable of meeting the system sa fety function, assuming a single failure. See Chapter 9 for details. See Chapter 8 for electrical system details. The Primary Component Cooling System is a closed system. During normal operation, it removes heat from various process system components, including letdown coolers and spent fuel pool coolers. This system also serves the residual heat removal and containment spray heat exchangers. Service water passes through the tube side of the primary component cooling water heat exchangers and transfers heat to the ultimate heat sink, which is either the Atlantic Ocean or a seismic Category 1 mechanical dr aft evaporative cooling tower. 3.1.4.16 Criterion 45 - Inspection of Cooling Water System The Cooling Water System shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system.

RESPONSE Important components of cooling water systems, such as pumps, strainers, heat exchangers and valves are located in accessible areas. These components have suitable manholes, handholes, inspection ports or other appropria te design and layout features to allow periodic inspection. In the case of buried piping, integrity wi ll be verified by pressure testing.

See Chapter 9 and 10 for details. 3.1.4.17 Criterion 46 - Testing of Cooling Water System The Cooling Water System shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of it components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant accidents, including operation of applicable portions of the Protection System and the transfer between normal and emergency power sources.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 29 RESPONSE The design provides for periodic testing of active components of the Cooling Water System for operability and functional performance as well as assuring the structural and leak-tight integrity of its components. Preoperational performance tests of the components are made in the manufacturer's shop. An initial system flow test demonstrates proper functioning of the system.

Thereafter, periodic tests ensure that components are functioning properly. Each active component of the Cooling Water System may be individua lly connected to the preferred power source at any time during reactor operation to demonstrate operability. Many active components are operating during normal operation, thereby demonstrating their availability. Remote operated valves may be exercised and actuation circuits tested. The automatic actuation circuitry, valves, and pump breakers also may be checked when integrated system tests are performed during a planned cooldown of the Reactor Coolant System. See Chapter 9 for details. See Chapter 8 for electrical system testing details. 3.1.5 Reactor Containment 3.1.5.1 Criterion 50 - Containment Design Basis The reactor containment structure, including access openings, penetrations, and the Containment Heat Removal System shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperatur e conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determinati on of the peak conditi ons, such as energy in steam generators and energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 30 RESPONSE The design of the containment is based on two containment design basis accidents. One assumes a double-ended rupture of the larg est reactor coolant pipe (LOCA); the other the rupture of a main steam line inside containment. The maximum calculated atmospheric pressure and temperature reached within the containment during the LOCA are 49.6 psig and 273 F; the maximum atmospheric pressure and temperature attained during a main steam line rupture is 36.1 psig and 364F. A containment design pressure of 52.0 psig has been selected to provide ample margin to allow for increased energy so urces. The peak liner temperature following a LOCA is calculated to be less than the design temperature of 271F. Although the containment atmospheric temperature following a MSLB is higher than that following a LOCA, the containment liner temperature will not exceed 271 F, since a lower heat transfer coefficient will result under the super-heated atmospheric condition during the MSLB. See Subsection 3.8.1 for containment loading combinations and Subsection 6.2.1 for design evaluation.

The containment electrical penetrations are designed so that the containment structure can, without exceeding the design leakage rate, accommodate the calculated pressure, temperature and other environmental conditions resulting from any loss-of-coolant accident. See Section 8.3 for discussion of containment elect rical penetrations and protecti ons of containment electrical penetrations. 3.1.5.2 Criterion 51 - Fracture Prevention of Containment Pressure Boundary The reactor containment boundary shall be designed with sufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a non-brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady-state, and transient stresses, and (3) size of flaws.

RESPONSE The reactor containment boundary is designed with sufficient margins to meet the requirements of Criterion 51.

All ferritic materials used for the fabrication of the containment liner and hatches are in accordance with ASME Boiler and Pressure Vessel Code,Section III, Divisions 2 and 1, respectively.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 31 For further details, refer to the following subsections: a. Concrete Containment 3.8.1 b. Steel Containment (Hatches) 3.8.2

c. Containment Functional Design 6.2.1 3.1.5.3 Criterion 52 - Capability for Co ntainment Leakage Rate Testing The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic inte grated leakage rate test ing can be conducted at containment design pressure.

RESPONSE The containment system is design ed and constructed and the necessary equipment is provided to permit periodic integrated leakage rate tests during plant lifetime, in accordance with reduced pressure test program requirements of Appendix J of 10 CFR 50.

Details concerning the conduct of periodic integrated leakage rate tests are presented in Chapter 6. 3.1.5.4 Criterion 53 - Provisions for Containment Testing and Inspection The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an a ppropriate surveillance program, and (3) periodic testing at containment design pre ssure of the leak-tightness of penetrations which have resilient seals and expansion bellows.

RESPONSE There are special provisions for conducting individual leakag e rate tests on applicable penetrations. Penetrations are visually inspected and pressure-tested for leak-tightness at periodic intervals, as required by Appendix J of 10 CFR 50.

Refer to Chapter 6 for details.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 32 3.1.5.5 Criterion 54 - Piping Systems Penetrating Containment Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities that reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

RESPONSE Piping systems penetrating primary reactor containment are provided with containment isolation valves, as described in responses to Criteria 55, 56 and 57.

Capability is provided for periodic testing of is olation valves during normal operation or during shutdown conditions in accordance with ASME Section XI, Article IWV.

Isolation valve leak test provisions are provided for complia nce with 10 CFR 50, Appendix J. See Chapter 6 for details. 3.1.5.6 Criterion 55 - Reactor Coolant Pressure Boundary Penetrating Containment Each line that is part of the re actor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis: a. One locked closed isolation valve insi de and one locked closed isolation valve outside containment; or b. One automatic isolation valve inside and one locked closed isolation valve outside containment; or c. One locked closed isolation valve inside and one automatic is olation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or d. One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 33 Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation va lves shall be designed to take the position that provides greater safety. Other appropriate requirements to minimize the pr obability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for in-service inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

RESPONSE Each line that is part of the re actor coolant pressure boundary that penetrates the containment is provided with isolation valves or barriers meeting this criterion. Instrument lines are designed in accordance with NRC Regulatory Guides 1.11, 1.141 and 1.151. Details and clarifications are provided in Subsection 6.2.4, 7.1.2.2a, and 7.3.1.1b. 3.1.5.7 Criterion 56 - Primary Containment Isolation Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis: a. One locked closed isolation valve insi de and one locked closed isolation valve outside containment; or b. One automatic isolation valve inside and one locked closed isolation valve outside containment; or c. One locked closed isolation valve inside and one automatic is olation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or d. One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 34 Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation va lves shall be designed to take the position that provides greater safety.

RESPONSE Each line that connects directly to the containment atmosphere and penetrates containment is provided with containment isolation valves or barriers meeting this criterion, except where it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable. Instrument lines are designed in accordance with NRC Regulatory Guides 1.11, 1.141, and 1.151.

Details are given in Subsections 6.2.4, 7.1.2.2a, and 7.3.1.1b. 3.1.5.8 Criterion 57 - Closed System Isolation Valves Each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside the containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

RESPONSE Each line that penetrates containment and is not connected directly to the containment atmosphere and is not part of the reactor cool ant pressure boundary has at least one isolation valve outside containment near the penetration, which meets this criterion. Details are provided in Subsection 6.2.4. 3.1.6 Fuel and Radioactivity Control 3.1.6.1 Criterion 60 - Control of Releases of Radioactive Materials to the Environment The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid e ffluents and to handle radioactiv e solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for re tention of gaseous and liquid e ffluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operation limitations upon the release of such effluents to the environment.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 35 RESPONSE Waste handling systems are incorporated in the faci lity design for processi ng and/or retention of normal operation radioactive wastes with appropriate controls and monitors to assure that releases do not exceed the guidelines of 10 CFR 20 and 10 CFR 50. The facility is also designed with provisions to prevent radioactivity release during accidents from exceeding limits of 10 CFR 100. The Radioactive Waste Processing System, the design criteria, and amounts of estimated releases of radioactive effluents to the environment are described in Chapter 11.

3.1.6.2 Criterion 61 - Fuel Storage and Ha ndling and Radioactivity Control The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering system, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

RESPONSE The Spent Fuel Pool and Cooling System, Fuel Handling System, Radioactive Waste Processing Systems, and other systems that contain radioactivity are designed to assure adequate safety under normal and postulated accident conditions.

Components are designed and located so that appropriate periodic inspection and testing may be performed.

All areas of the plant are designed with suitable shielding for radiation protection based on anticipated radiation dose rates and o ccupancy as discussed in Chapter 12. Individual components that cont ain significant radioactivity are in confined areas adequately ventilated through appropriate filtering systems. The Spent Fuel Cooling System provides cooling to remove residual heat from the fuel stored in the spent fuel pool. The system is designed with redundancy and testability to assure continued heat removal. The spent fuel cooling system is described in Chapter 9.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 36 The spent fuel pool is designed so that no postulated accident could cause an excessive loss of coolant inventory. 3.1.6.3 Criterion 62 - Prevention of Critica lity in Fuel Storage and Handling Criticality in the Fuel Storage and Handling System shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

RESPONSE Criticality in fuel storage areas is prevented by both physical separation of fuel assemblies and the presence of borated water in the fuel storage pool. The fuel storage racks are constructed so that the fuel assemblies may be inserted in prescribed locations only. These racks are designed with neutron absorbing material as an integral rack component to ensure sub-criticality even if assemblies are immersed in unborated water. Criticality prevention and criticality considerations are discussed in Sections 9.1 and 4.3, respectively.

3.1.6.4 Criterion 63 - Monitoring Fuel and Waste Storage Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.

RESPONSE Monitoring systems are provided to alarm when excessive temperatures or low water level occurs in the spent fuel pool. Appropriate safety actions are initiated by operator action. Radiation monitors and alarms are provided as required to warn personnel of impending excessive levels of radiation. The ventilation system in the Fuel Storage Building is operated in the fuel handling mode anytime irradiated fuel not in a cask in handled. In this operating mode, the building is maintained at a negative pressure and all exhaust air is filtered by charcoal filters prior to discharging the air to the atmosphere via the plant unit vent. In the unlikely event of a fuel handling accident, the filtration system is already operational and av ailable to filter exhaust air from the building. See Chapter 9 for details. The Radiation Monitoring System is described in Chapter 12.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS Conformance to NRC General Design Criteria Revision 12 Section 3.1 Page 37 3.1.6.5 Criterion 64 - Monitoring Radioactivity Releases Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released for normal operations, including anticipated operational occurrences, and from postulated accidents.

RESPONSE The containment atmosphere is continually monitored during normal and transient station operations, using the containment air particulate and radio-gas radiation monitors. In accident conditions, samples of the containment atmosphere provide data on existi ng airborne radioactive concentrations within the containment. Radioactivity levels contained within the facility gaseous and liquid effluent discharge paths and in the station environs are continually monitored during normal and accident conditions by the station radiation monitoring as described in Chapter 11 and 12. In addition to the installed detection system, an environment radiation surveillance program monitors the exposure pathways to Man within a 5-mile radius from the station site.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 1 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS

3.2.1 Seismic

Classification Seabrook Station structures, systems and components important to safety, as well as their foundations and supports, have been designed to withstand the effects of an Operating Basis Earthquake (OBE) and a Safe Shutdown Earthquake (SSE), and are thus designated as seismic Category I. These plant features are those necessary to assure: a. The integrity of the reactor coolant pressure boundary, b. The capability to shut down the reactor and maintain it in a safe shutdown condition, or c. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comp arable to the guideline exposures of 10 CFR 100. Those structures, systems and components classified as seismic Category I are listed in Table 3.2-1 and Table 3.2-2. These structures, systems and components are designed to withstand seismic loads as discussed in Section 3.7.

Seismic Category I structures are sufficiently isolated or protected for non-seismic Category I structures to ensure that their integrity is maintained at all times. None of the plant structures is classified as partially seismic Category I, with the exception of the Circulating Water Pumphouse and the Waste Processing Building, which are described in Section 3.8. Several non-seismic Category I structures are designed against collapse onto seismic Category I structures due to SSE loadings. These are discussed in Subsection 3.7(B).2.

Where only portions of systems are classified as seismic Category I, the boundaries of the seismic Category I portions of the system are shown on the ductwork, piping and instrumentation diagrams in the referenced sections of the Updated FSAR. The seismic Category I portion of these systems extends to the first restraint beyond th e isolation valves that isolate the part which is Category I from the nonseismic portion of the system. Equipment and components that are not classified as seismic Category I, and whose collapse or failure could result in the loss of safety function of seismic Category I structures, systems or components, are checked to confirm their structur al integrity against collapse or failure due to SSE loadings.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 2 Where the collapse on non-Category I components would adversely affect the performance of Category I systems or components, they have been supported to withstand seismic loadings or isolated from the Category I systems or compone nts by boundary restraints. The locations of the boundary restraints are shown on the pipe support location drawings and the pipe support detail drawings. Where the collapse of non-Category I HVAC system components or ductwork would damage the other Category I system or components in the vi cinity, they have been supported to withstand seismic loadings.

The seismic classifications presented in Table 3.2-1 and Table 3.2-2 are consistent with the recommendations of NRC Regulatory Guide 1.29.

3.2.2 Quality

Group Classification The system quality group classification applies to fluid systems, or portions of fluid systems, which are directly depended upon to: a. Prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary b. Permit shutdown of the reactor and maintenance in the safe shutdown condition c. Contain radioactive material. Fluid system components important to safety are classified in accordance with the ANSI N18.2a-1975, "Revision and Addendum to Nuclear Sa fety Criteria for the Design of Stationary Pressurized Water Reactor Plants." This classification system is compatible with the requirements of NRC Regulatory Guide 1.26, and is submitted as an alternate acceptable method of meeting the intent of Regulatory Guide 1.26.

3.2.2.1 Safety Class Definitions Components are classified as Safety Class 1, Safety Class 2, Safety Class 3, and nonnuclear-safety in accordance with their importance to nuclear safety. This importance, as established by the assigned safety class, is applied in the design, materials, manufacture or fabrication, assembly, erection, construction, and operation. A single system may have components in more than one safety class.

The definitions of safety classes listed apply to fluid pressure boundary components and the reactor containment. Supports that have a nuclear safety function are of the same safety class as the components that they support.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 3 a. Safety Class 1 Safety Class 1 applies to components whose failure could cause a Condition III or Condition IV loss of reactor coolant. Condition III occurrences include incidents any one of which may occur during the lifetime of a plant. Condition IV occurrences are faults that are not exp ected to occur but are postulated because their consequences would include the potential for release of significant amounts of radioactivity. Condition IV faults are the most drastic which must be designed against, and thus represent the limiting design case.

b. Safety Class 2 Safety Class 2 applies to the reactor containment and to the following components: 1. Components of the reactor coolan t pressure boundary not in Safety Class 1. 2. Components of safety systems that are necessary to remove heat directly from the reactor containment, to circ ulate reactor coolant for any safety system purpose, to control within the reactor containment radioactivity released, or to control hydrogen in the reactor containment. A safety system (in this context) is any system that is necessary to shut down the reactor, cool the core or cool another safety system or the reactor containment (after an accident), or it is any system that contains, controls, or reduces radioactivity released in an accident. Only those portions of a system that are designed primarily to accomplish one of those functions, or the failure of which could prevent accomplishing one of those

functions, are included.

c. Safety Class 3 Safety Class 3 is applied to those components not in Safety Class 1 or Safety Class 2, the failure of which would result in release to the environment of radioactive gases normally required to be held for decay, or that are necessary to: 1. Provide or support any safety system function,
2. Control outside the reactor containmen t airborne radioactivity released in an accident, or S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS Classification of Structures, Components, and Systems Revision 8 Section 3.2 Page 4 3. Remove decay heat from spent fuel.
d. Nonnuclear Safety Nonnuclear safety applies to portions of the nuclear power plant not covered by Safety Classes 1, 2, or 3 that can influence safe, normal operation or that may contain radioactive fluids. Design of nonnuclear safety components shall be to applicable industry codes and standards.

3.2.2.2 Component Classification and Code Requirements Fluid system component safety class designations are listed in Table 3.2-2. Table 3.2-2 also indicates industry codes and standards applicable to pressure-retaining components and associated safety systems, and provides other pertinent information. The safety classification and design standards for instrument sensi ng lines follow the guidelines of Regulatory Guide 1.151 with exceptions as listed in Subsection 7.1.2.12. The safety class designation boundaries of safety-related systems are shown on the piping and instrumentation diagrams in the referenced sections of the Updated FSAR. HVAC system component safety class designations are listed in Table 3.2-4. Table 3.2-4 also indicates design criteria and guidelines used for ductwork and supports. The safety class designation boundaries of safety-related systems are shown on air flow diagrams and ductwork drawings.

Quality standards, as specified by Code classifications, are correlated to safety class designations in Table 3.2-3. Quality assurance programs applied to safety-related systems and components, and developed to meet the intent of 10 CFR 50 Appendix B, are presented in Chapter 17.

S EABROOK D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND Revision 8 SYSTEMS S TATION Section 3.3 UFSAR Wind and Tornado Loadings Page 1 3.3 WIND AND TORNADO LOADINGS 3.3.1 Wind Loadings The design of seismic Category I structures for normal wind loading is based on two criteria: (1) ASCE Paper No. 3269, "Wind Forces on Structures," Reference 1 (Subsection 3.3.3) and (2) ANSI A58.1-1972, "Building Code Requirements for Minimum Design Loads in Buildings and Other Structures," Reference 2 (Subsection 3.3.3). The ASCE paper is used to derive the wind loading for the Containment Enclosure Building, while the ANSI guideline is used for wind loading on all other seismic Category I structures.

3.3.1.1 Design Wind Velocity The design wind velocity at 30 feet above ground for the 100 year period of recurrence is 110 mph (See Subsection 2.3.1). a. ASCE - The vertical wind velocity profile is interpolated from Table 1(b) of Reference 1 (Subsection 3.3.3). A gust fact or of 1.1 is used in deriving wind loading based on guidelines presented in Reference 1 (Subsection 3.3.3). b. ANSI - The vertical velocity profile a nd applicable gust factors are included in the velocity pressure profiles discussed in Subsection 3.3.1.2b.

3.3.1.2 Determination of Applied Forces Wind loads are applied as uniform static loads on the horizontal and vertical projected areas of the structure walls and roof. Shielding effects provided by other structures are neglected unless a portion of the exposed surface is immediately adjacen t to another structure. Only dead loads are considered in resisting uplift. a. ASCE - The design wind velocities ar e converted to wind pressures by the following formula: C p 2GV0.00256p where: p = design wind pressure (psf) G = gust factor V = wind velocity (mph)

C p = pressure coefficient (Table 4, Reference 1)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 2 The vertical profile of wind pressure on the Containment Enclosure Building is summarized in Table 3.3-1. The distributi on of effective pressure coefficients, (C p), for cylindrical and spherical structures is shown on Figure 3.3-1. b. ANSI - The effective velocity pressure vertical profile, including gust factors for structures (q F), parts of structures (q p), and internal pressure (q M) are taken from Tables 5, 6 and 12 (Exposure C) respectively of Reference 2. The effective velocity pressures are summarized in Ta ble 3.3-2. The average pressure acting on a structural element is as follows: p = q C p - q M C pi where:

p = average wind pressure (psf) q = q F or q p whichever is appropriate C p = external pressure coefficient q M = internal velocity pressure C pi = internal pressure coefficient Load combinations involving normal wind (discussed in Subsections 3.8.4.3 and 3.8.5.3) do not control the design of seismic Category I structures.

3.3.2 Tornado

Loadings 3.3.2.1 Applicable Design Parameters

a. Tornado wind loads on seismic Category I structures are based on a single vortex tornado model which results in a maximum tangential velocity of 290 mph and a maximum translational velocity of 70 mph. The total maximum wind velocity is 360 mph assumed constant with respect to height. The pressure differential is 3 psi at a rate of 2 psi/second. See Subsection 2.3.1.2 for a discussion of the design basis tornado parameters.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 3 b. Protection of equipment and systems locat ed in the "closed" structures listed in paragraph c. below, is accomplished by using protective barriers consisting of dampers complete with the necessary accessories and instrumentation. These dampers are provided at inlet openings located in the exterior walls of the structures which provide outdoor air for the plant HVAC systems. These dampers are also provided at the openings in the exterior walls of the structures, through which air is exhausted from the various plant HVAC systems to the outside atmosphere. The dampers automatically close to avoid damage which could be

caused by tornado induced depressurization within the structures to the systems and components located inside the structures. Ductwork and backdraft dampers located internal to the tornado dampers were evaluated for the positive tornado wind pressure loads considering the tornado dampers open. Seismic Category I systems and components which are located in "open" structures are protected from tornado wi nd pressure loads and tornado-generated missiles. Such equipment and interior st ructures have been shown to be capable of performing their design function when subjected to tornado depressurization effects. c. Of the seismic Category I structures described in Subsection 3.8.4.1, the following structures are protected from tornado depressurization effects and are considered "closed" structures. 1. Containment Enclosure Ventilation Area (including Containment)

2. Control Building
3. Electrical Tunnels
4. Emergency Feedwater Pump Building
5. Primary Auxiliary Building (including equipment vault areas)
6. Service Water Pumphouse (Switchgear Room Only)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 4 3.3.2.2 Determination of Forces on Structures

a. The total wind velocity is transformed into the total tornado wind load on seismic Category I structures by means of either the ASCE Paper 3269, "Wind Forces on Structures" (Reference 1), or the ANSI Standard A58.1-1972, "Building Code Requirements for Minimum Design Loads in Building and Other Structures" (Reference 2). The ASCE Paper is used for the Containment Enclosure Building and the ANSI Standard is used for all other seismic Category I structures. Velocity pressure is assumed to be constant with height. Maximum velocity pressure is based on the maximum tornado wind velocity and is assumed to occur at the radius of the tornado funnel at which the maximum velocity occurs. The maximum velocity pressure, qmax, is given by the following formula:

qmax = 0.00256V 2 where, V = total tornado wind velocity - 360 mph therefore, qmax = 332 psf q avg = qmax C s where, q avg = Average velocity pressure C s = Size factor depending on L (See Figure 3.3-2) L = Load distribution length equal to: 1. the plan distance perpendicular to the wind over which the wind load can be distributed, or 2. the mean horizontal extent of the tributary area perpendicular to the direction of wind. The total pressure (W w) acting on a structure or structural element is equal to the external tornado wind pressure (W w1) plus the internal wind pressure (W w2). Calculation of the two pressure components (W w1 and W w2) for the two controlling criteria (ASCE and ANSI) is shown in Table 3.3-3.

S EABROOK D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND Revision 8 SYSTEMS S TATION Section 3.3 UFSAR Wind and Tornado Loadings Page 5 Variation of tangential velocity with radial distance from the tornado core is determined as follows:

rr0formax V r r V m t m t r r rformax V r r V 75 m t r m t where V t = tangential velocity at radius r Vt max = maximum tangential velocity (290 mph) r = radius from centerline of tornado r m = radius of maximum tangential velocity (150 ft) r 75 = radius at which tangential velocity equals 75 mph (580 ft) b. Venting of structures is not adopted as a means of transforming the tornado-generated differential pressure into an effective reduced pressure. The exterior walls and roof slabs of seismic Category I structures are subjected to the full 3 psi pressure drop. c. The postulated horizontal and vertical tornado-generated missiles, and procedures used to design the structures or barriers against the effects of these missiles are described in detail in Subsections 3.5.1.4 and 3.5.3, respectively. d. The effective wind loadings are combined in the following manner in order to determine the most adverse tornado effect on seismic Category I structures. (I) W t = W w1 + W w2 (II) W t = W p (III) W t = W m (IV) W t = W w1 + 0.5W p (V) W t = W w1 + W w2 + W m (VI) W t = W w1 + 0.5W p + W m S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 6 where, W t = Total tornado load W w1 = Tornado wind external pressure load W w2 = Tornado wind internal pressure load W p = Tornado differential pressure drop = 3 psi (432 psf)

W m = Tornado missile load The most severe W t of the above combinations is used in the seismic Category I load combinations discussed in Subsections 3.8.1 and 3.8.4.

3.3.2.3 Effect of Failure of Structures or Components Not Designed for Tornado Loads Any non-Category I structure, which as a re sult of a catastrophic failure could fall upon a seismic Category I structure and, thereby, effect th e safety function of the structure, is designed for tornado wind in the following manner: a. Exterior siding, nonstructural walls and partitions, and the roof slab are considered expendable and can be permitted to fail during a tornado. The failure of these elements does not generate missiles of greater severity than those listed

in Subsection 3.5.1.4. b. The structural frame is checked for tornado wind pressure by assuming that the wind load acts on the wall fully sided, or on one-third of the siding area and the remainder of the exposed steel frame including girts, crane girders, etc.,

whichever results in the most severe load

s. The resulting tornado wind loading is used in the load combination for seismic Category I structures (other than containment), and the stresses checked against the appropriate acceptance criteria.

(See Subsection 3.8.4.) The nonseismic Category I structures which are designed for tornado wind using the procedure outlined above are summarized in Table 3.3-4.

The large roll-up steel door located in the east wall of the seismic Category I Fuel Storage Building is not designed for tornado loadings. Its failure will not result in any adverse consequences affecting the safety function of th e Fuel Storage Building or adjoining structures.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Wind and Tornado Loadings Revision 8 Section 3.3 Page 7 3.3.3 References

1. ASCE Paper No. 3269, "Wind Forces on Structures," Transactions, Vol. 126, Part II, 1961, P1124. 2. ANSI A58.1-1972, "Building Code Requirements for Minimum Design Loads in Buildings and Others Structures," Am erican National Standards Institute, New York, 1972.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Water Level (Flood) Design Revision 16 Section 3.4 Page 1 3.4 WATER LEVEL (FLOOD) DESIGN

3.4.1 Flood

Protection 3.4.1.1 Flood Protection Measures for Seismic Category I Structures All safety

-related systems and components, as identified in Table 3.2-1 and Table 3.2-2, are protected against floods. These systems and components are protected by the structures which house them, and/or by being located above a maximum water level not exceeding 21 feet MSL, postulated to result during the combined PMH

-SPS event.

Seismic Category I structures house safety

-related equipment, and are listed in Table 3.2-1. These structures are designed to withstand a depth of still water of not exceeding 21 feet MSL. The locations of the walls of those structures which could be subject to wave runup to elevation of 21.8 feet MSL are shown in Figure 2.4-21. The only access openings in any exterior wall that are below the design flood level are the rolling steel door in the Fuel Storage Building (west wall) and a personnel emergency exit (east wall), located at elevation 20 feet 6 inches MSL (See Figure 1.2-16), and the double doors into the entrance vestibule of the Equipment Vault section of the Primary Auxiliary Building, located at elevation 20 feet 8 inches MSL (See Figure 1.2-13). Flood protection for the Fuel Storage Building is provided by a curb at elevation 21.5 feet MSL located on column line 3. This door is closed during normal plant operation, thus providing the same protection against wave run

-up as the other vertical building walls. The floor of the vestibule into the Equipment Vault section of the Primary Auxiliary Building is sloped up 4 inches so that the high point in the floor is at elevation 21 feet MSL. To minimize potential in

-leakage from such phenomena as minute cracks in structure walls or leakage waterstops, all below

-grade safety

-related structures, other than the pumphouse

, cooling tower, electrical duct banks and manholes are waterproofed on the exterior face. Such cracks will be minimal because of the structures being heavily reinforced due to the various design criteria. Typical details of waterproofing for penetrations are shown in Figure 3.4-1. In addition, sump pumps are provided in all seismic Category I structures other than Category I manholes where seepage that could occur could affect safety

-related equipment. All structures except the Control Building, which is above grade, are protected in this manner. The pipe chases below the Control Building drain into the Emergency Feedwater Pump Building, which is protected by sump pumps.

Because all safety

-related structures, systems and components are provided with protection against floods, as described in the preceding paragraphs, it will not be necessary to bring the reactor to a cold shutdown for the most severe flood anticipated at the plant.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Water Level (Flood) Design Revision 16 Section 3.4 Page 2 Protection against possible flooding from liquid carrying systems due to pipe rupture or fire protection activities is discussed in the following sections: Section 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Subsection

9.3.3 Equipment

and Floor Drainage System Subsection 9.5.1 Fire Protection System 3.4.1.2 Plant Dewatering System Seabrook Station was not originally designed with a dewatering system, because it was believed that the in

-leakage prevention methods described in 3.4.1.1 would be adequate to prevent water ingress. Over the years, it has become evident that the mitigation methods were not completely effective at preventing in

-leakage. The presence of groundwater in

-leakage is a plant concern from the perspectives of:

Housekeeping (i.e., maintaining walking surfaces in lower elevations of building structures, below grade, dry) is difficult because the in

-Leakage is continuous In-leakage occurs in some areas where it could spread (tritium) contamination (i.e., a small wet area becomes a larger wet area)

In-leakage allows communication of certain plant drainage systems (e.g., the spent fuel pool leak off system) with the soil and groundwater beneath the buildings.

A plant dewatering system has been installed which can further mitigate in

-leakage of groundwater in the lower elevations of the plant. The purpose is to routinely pump water from beneath the plant structures, to reduce the static hydraulic head outside the building concrete and reduce the in

-leakage. This allows the original mitigative measures to function properly. A pump is installed in the existing well at (+)7' elevation of the PAB. This pump discharges the water to the roof drain system, which then flows to the storm drain system and out to circulating water for discharge.

Routine monitoring of this flowpath is performed per station operating procedures.

Existing pipe penetrations located in EF103/EFST1 have been utilized as a groundwater low point. These penetrations have been directed to a nearby sump (1

-DF-5). This sump discharges to the existing plant storm drainage system.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Water Level (Flood) Design Revision 16 Section 3.4 Page 3 A pump is installed in the RHR Vault "B" stairwell at (-)61' elevation of the Equipment Vault. This pump discharges the water to the roof drain system, which then flows to the storm drain system and out to circulating water for discharge. Routine monitoring of this flow path is performed per station operating procedures.

A pump is installed in the containment annulus at

-32' elevation. This pump discharges the ground water in the containment annulus to the roof drain system. The connection to the drain system is installed at 240 degrees azimuth of the containment annulus. Routine monitoring of this flow path is performed per station operating procedures.

A ground water collection tank and pump are installed in the "B" Electrical Tunnel, West stairwell at (-) 20'-0" elevation. The pump discharges the water to the turbine building roof drain system, which then flows to the storm drain system and out to circulating water for discharge via the outfall. Routine monitoring of this flow path is performed per station operating procedure. A 50.59 Applicability and Screening are performed to support this change.

3.4.2 Analytical

and Test Procedures The methods and procedures by which the static and dynamic effects of the design basis flood conditions or design basis groundwater conditions identified in Section 2.4 are applied to safety-related structures, systems and components are described in Subsections 3.8.1, 3.8.4 and 3.8.5. In the design of seismic Category I structures, the groundwater is assumed to be at 20 feet MSL, the plant grade. As discussed in Subsection 3.4.1, the design basis flood could cause a depth of stillwater on the plant grade of not exceeding 1 foot, increasing the top of the hydrostatic head to elevation not exceeding 21 feet MSL. Dynamic effects of the design basis flood were considered, but found to be negligible. The maximum depth of stillwater is not exceeding one foot above plant grade, and the maximum wave runup in local regions is 1.8 feet above plant grade. Any dynamic effects produced by these occurrences were evaluated and found to be negligible and, due to the relatively large masses of the reinforced concrete structures, can be neglected.

The coincident wind loadings used in the design of the structures are those associated with the Probable Maximum Hurricane (PMH), during which the design basis flood occurs. These wind loading are discussed in Subsection 3.3.1. The static and dynamic effects on foundations due to the design basis flood are considered in the design of all seismic Category I structures. Dynamic forces due to wave runup are negligible compared to the forces that are considered because of other design criteria.

All seismic Category I structures are designed to prevent uplift or overturning, considering the effects of buoyant forces along with other forces (see Subsection 3.8.5).

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 1 3.5 MISSILE PROTECTION Structures, shields and barriers are provided as pr otection against the effects of both internally and externally generated missiles, an exception being the turbine-generated missiles discussed in Subsection 3.5.1.3. Structures that have been analyzed for missile damage protection capabilities are those that: House or service safety-related systems and components required for the safe shutdown of the reactor and its maintenance in a safe shutdown condition; If damaged, could result in a significant uncontrolled release of radioactivity. 3.5.1 Missile Selection and Description 3.5.1.1 Internally-Generated Missiles (Outside Containment)

The principal design basis for missiles generated outside containment, but internal to the plant site, is that such missiles shall not cause loss of function of any design f eature provided for either continued safe operation or shutdown of the reactor during operating conditions, operational transients, and postulated accident conditions associated with the effects of missile formation.

Two general sources of internally generated missiles outside containment which have been considered are: a. Those due to rotating component failures

b. Those due to pressurized component failures. Components evaluated for possible missile generation outside containment, and discussed below, are valves in high pressure systems, temperature sensing element wells, pumps, and motor generator sets.

Catastrophic failure of major components, e.g., pum ps, heat exchangers, etc., is not considered credible because of: (a) material characteristic s, inspections, quality control during fabrication, erection, and operation; (b) the maximum no-load speed of the components is the operating speed of their motors; (c) fans are tested to 125 percent of operating speed; and (d) where necessary internally generated missile analysis will be conducted to verify that no such missiles can degrade the adjoining safety-related equipment.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 2 a. Valves 1. Valve stems are not considered to be credible sources of missiles, since all valves in high energy lines have stems with a back seat that effectively eliminates the possibility of valve stems being ejected even if the stem threads fail. Analysis has indicated that the back seat or the upset end will

not penetrate the bonnet. Additional in terference is encountered with air and motor-operated valves. 2. Valves with a nominal diameter larg er than 2 inches have been designed against bonnet-body connection failure and subsequent bonnet ejection by: (a) Using the design practice of ASME Section III which limits the allowable stress of bolting material to 25 percent of minimum yield strength (b) Using the design practice of ASME Section III for flange design (c) Controlling the load during the bonnet-body connection stud tightening process. 3. Valves with nominal diameter of 2 inches or smaller are forged and have screwed or bolted bonnets. Bonnet fastening systems are designed to code requirements. The pressure-containing parts are designed per criteria

established by ASME Section III code specification. 4. The pressure-containing parts of va lves, including the flanges and studs, are designed to the requirements of the ASME Code for safety class valves and ANSI B16.34 and B16.5 for non-nuclear safety class valves. 5. The proper stud torquing and tightening procedures, and the use of torque wrenches, with indication of the applied torque or stud tension, limit the stress of the studs to the allowable limits established in the ASME Code. This stress level is far below the material yield. 6. The complete valves are hydrotested per code requirements. 7. Suitable nondestructive tests are performed, as appropriate, in accordance with code requirements. It is, therefore, concluded that there are no credible sources of missiles associated with valves located outside containment.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 3 b. Temperature Sensing Element Wells Temperature sensing element wells in safety-related system piping outside containment are designed, manufactured and welded to the same codes as the piping in which they are located and are, therefore, also not considered credible missiles. Where possible, the sensing element well is located so that, if in the remote case it did become a missile, it would impact directly on a nearby

reinforced concrete wall, floor or ceiling. These barriers are designed to withstand such missiles. Where required, a local missile barrier will be designed and installed in individual cases. c. Pumps 1. Motor-Driven Pumps located outside containment have been evaluated for missiles associated with overspeed failure. The maximum no-load speed for these pumps is equivalent to the operating speed of their motors. Consequently, no pipe break or single failure in the suction line would increase pump speed over that of the no-load condition. Furthermore, there are no pipe break plus single failure combinations which could result in a significant increase in pump suction or discharge head. Therefore, no overspeed is expected, and missiles associated with pumps outside containment are not

credible.

2. Turbine-Driven The turbine-driven Emergency Feedwater (EFW) pump is identical in size and design to the motor-driven EFW pump, and operates at the same speed. The turbine-driven EFW pump is not considered to be a credible missile source for the same reasons stated above for the motor-driven pump. The turbine unit is equipped with both a speed-limiting governor and an overspeed limiting trip. The Speed Governing System is designed to

assure rapid controlled acceleration w ithout overspeeding. The overspeed governor consists of a mechanical pin-type device which trips shut the turbine steam inlet valve at 125 percent overspeed. Repeatability accuracy of this trip is within +/- 50 rpm.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 4 The turbine itself is of a solid wheel, single stage design and is not considered a credible missile source. This unit has been designed to start or operate on 100 percent water as we ll as being able to withstand the severe punishment of intermittent water slugs. During testing of similar type units, water slugs were injected into the turbine while it was operating normally. These water slugs ranged from 50 to 600 gallons.

Following these tests, detailed examinations confirmed that the turbine sustained no wear deformation or damage. Further assurance of the integrity of the turbine drive unit is provided by the Quality Assurance program of the vendor. During manufacturing, the vendor follows the intent of ASME Section III standards, Appendix B to 10 CFR 50 (Quality Assurance Criteria Requirements of the Code of Federal Regulations) and ANSI 45.2. Certified material is used on all major components so that complete traceability is possible. Welders are qualified to ASME Section IX st andards and non-destructive test personnel are qualified to SNT standards.

The integrity of every turbine casing is confirmed by 100 percent mag-particle testing. Every high pressure component is subjected to a thorough X-ray analysis. Every shaft, as well as every wheel is ultrasonically tested. Additionally, conversations with the turbine manufact urer have indicated that for the turbine wheel to separate, speeds in excess of 14,000 RPM would be necessary. This is approximately a 400 percent overspeed for this unit. The motor-driven EFW pump is oriented perpendicular to the turbine-driven EFW pump so that in the unlikely event that pump or turbine missiles are generated the other pump will not be affected.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 5 d. Motor-Generator Sets The fabrication specifications of the mo tor-generator set flywheels control the material to meet ASTM A533-7D, Grade B, Class 1, with inspections per MIL-I-45208A and flame cutting and machin ing operations governed to prevent flaws in the material. Nondestructive testing for nilductility (ASTM E208), Charpy V-notch (ASTM A593), ultrasonic (ASTM A577 and A578) and magnetic particles (ASME Section III, NB2545) is performed on each flywheel material lot. In addition to these requirements, stress calculations are performed

consistent with guidelines of the ASME Code Section III, Appendix A, to show that the combined primary stresses due to centrifugal forces and the shaft interference fit will not exceed one-third of yield strength at normal operating speeds (1800 rpm), and will not exceed tw o-thirds of the yield strength at 25 percent overspeed. However, no oversp eed is expected because the flywheel weighs approximately 1300 pounds and has dimensions of 35.26 inches in diameter by 4.76 inches wide. The flywheel mounted on the generator shaft, which is directly coupled to the motor shaft, is driven by a 200 hp, 1800 rpm induction motor. The torque developed by the motor is insufficient for overspeed. Therefore, there are no credible missiles from the motor-generator sets. e. Accumulators Accumulators not designed to ASME Section III are not considered a source of credible missiles since they are designed in accordance with ASME Section VIII. Various subsections of ASME Section VIII, e.g. UG-22, UG-23 and UCS-66 delineate requirements for impact testing necessary to prevent brittle fracture.

The Seabrook tankage (accumulators) designed to Section VIII have either appropriate impact test, or material-operating temperatures that preclude brittle fracture. Thus, missiles from this source are not considered credible. Various measures, such as separation of redundant safety-related equipment, have been employed to assure that essential equipment is protected against postulated missiles.

Structures, shields and barriers that are designed to withstand missile effects are tabulated in Table 3.5-1.

A tabulation of all safety-related structures, systems, and components is presented in Table 3.2-1 and Table 3.2-2. General arrangement and section detail drawings are located in Section 1.2.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 6 3.5.1.2 Internally Generated Missiles (Inside Containment)

The principal design bases for protection against the effects of missiles generated within the reactor containment, coincident with a loss-of-coolant accident, are that: a. Missiles generated in one reactor coolant loop shall not propagate damage to unaffected loops. b. The function of any redundant engineered safety feature shall not be lost. c. Containment integrity shall be maintained. Equipment located inside containment has been evaluated for potential missile sources. As a result of this review, the following information is provided concerning potential missile sources and systems which require protection from internally generated missiles inside containment.

Catastrophic failure of the reactor vessel, steam generators, pressurizer, reactor coolant pump casings, and piping leading to the generation of missiles, is not considered credible because of material characteristics, inspections, and quality control during fabrication and erection of the particular component.

All equipment, piping and supports in containment are either secured with seismic anchors to withstand seismic disturbances or are isolated by their respective locations from safety-related equipment. Hence, there is no credible source of gravitational missiles. The reactor coolant pump flywheel is not considered a source of missiles for the reasons discussed in Subsection 5.4.1. Nuclear steam supply system components which are considered to have the potential for missile generation inside the reactor containment are identified as follows: Control rod drive mechanism housing plug, drive shaft, and the drive shaft and drive mechanism latched together Certain valves Temperature and pressure sensor assemblies Pressurizer heaters.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 7 a. Control Rod Drive Mechanism Components Gross failure of a control rod drive mechanism housing, sufficient to allow a control rod to be rapidly ejected from the core, is not considered credible for the following reasons: Control rod drive mechanisms are shop hydrotested at 4100 +/-75 psi.

Control rod drive mechanism housings are individually hydrotested to 3107 psi after they are installed on the reactor vessel to the head adapters, and checked again during the hydrotest of the completed Reactor Coolant System. Control rod drive mechanism housings are made of Type 304 stainless steel. This material exhibits excellent notch toughness at all temperatures that will be encountered. It is, however, postulated that the top plug on the control rod drive mechanism can become loose and be forced upward by the water jet. The following sequence of events is assumed: 1. The drive shaft and control rod cluste r are forced out of the core by the differential pressure of 2500 psi across the drive shaft (the drive shaft and control rod cluster, latched together, are assumed fully inserted when the accident starts); 2. After approximately 12 feet of travel, the rod cluster control spider hits the underside of the upper support plate; 3. Upon impact, the flexure arms in the coupling joining the drive shaft and control cluster fracture, completely freeing the drive shaft from the control

rod cluster; 4. The control cluster will be completely stopped by the upper support plate; however, the drive shaft can continue to be accelerated upward to hit the missile shield.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 8 The missile shield is integrated into the simplified head assembly and consists of a reinforced two-inch thick steel plate with ventilation holes located approximately 1.5 feet above the rod travel housings. The characteristics of the postulated control rod drive missiles are given in Table 3.5-2. Penetration evaluation was performed as required by the Standard Review Plan, Section 3.5.3, Rev. 1, 1980 (NUREG 800).

b. Valves Valves within the reactor coolant pressure boundary have been examined to identify potential missiles. It was conc luded that there are no credible valve failures that could result in missile formation, since valve design features effectively preclude the ejection of valve stems. Valves are designed against bonnet-to-body connection failure and subsequent bonnet ejection by: 1. Compliance with the ASME Boiler and Pressure Vessel Code,Section III
2. Control of load during tightening of bonnet-to-body joints. For the special case of those valves lo cated on the top of the pressurizer that extend above the operating deck, certain missiles, although incredible, are postulated. Protection has, therefore, been provided due to potential damage to the containment liner, engineered safeguards pipes and components located outside the secondary compartments. Missile characteristics of the bonnets of the valves in the region where the pressurizer extends above the operating deck are given in Table 3.5-3. c. Temperature and Pressure Sensing Assemblies The only credible source of jet-propelled missiles originating from reactor coolant piping and piping with systems connected to the Reactor Coolant System is that represented by the temperature and pressure sensor assemblies. The resistance temperature sensor assemblies can be of two types ("with well" and "without well"). Two rupture locations have b een postulated: one around the welding (or thread) between the temperature element assembly and the boss for the without-well element, and the welding (o r thread) between the well and the boss for the with-well element.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 9 A temperature sensor is installed on the reactor coolant pumps close to the radial bearing assembly. A hole is drilled in the gasket and sealed on the internal end of a steel plate. In evaluating missile potential, it is assumed that this plate could break, and the pipe plug on the external end of the hole could become a missile. In addition, it is assumed that the welding between the instrumentation well and the pressurizer wall could fail, and the well and sensor assembly could become a jet-propelled missile. Where possible, the sensing element well is located so that if it were to become a missile, it would impact on a nearby concre te wall, floor or ceiling. The weights and impact velocities for such postulated missiles are shown in Table 3.5-4. A

10 - degree expansion half-angle water jet has been assumed. For the weights and velocities listed, they are not considered capable of generating secondary missiles upon impact with the above concrete ba rriers. These barriers are designed to withstand such missiles. The missile characteristics of the piping pressure element assemblies are less severe than those of Table 3.5-4. The missile characteristics of the reactor coolant pump temperature sensor, the instrumentation well of the pressurizer, a nd the pressurizer heaters are given in Table 3.5-5. A 10-degree expansion half-angle water jet has been assumed. 3.5.1.3 Turbine Missiles

a. Turbine Placement and Orientation The placement and orientation of the turbine generators and potential missile ejection zones of +/-25 degrees with respect to the low-pressure-end turbine wheels

for the turbine unit is shown in Figure 3.5-1. Plan and elevation views of the Turbine Building are given in Figure 1.2-37 Figure 1.2-38, Figure 1.2-39, Figure 1.2-40, Figure 1.2-41, Figure 1.2-42, Figure 1.2-43, Figure 1.2-44 and Figure 1.2-45. Plan and layout drawings of other plant structures are also presented in Section 1.2.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 10 b. Missile Identification and Characteristics Experience and calculation indicate that in the improbable event of a rotor fracture, the substantial fragments of the high-pressure turbine sections and generator rotors would be contained within their respective casings (Reference 1). Accordingly, missiles analyzed in this section will be limited to postulated low-pressure turbine missiles. 1. Low-Pressure Turbine Missile Characteristics It is considered that the failure of any of the shrunk-on wheels in the seven low pressure stages of each half of the three turbine units at both design overspeed and runaway speed can lead to the generation of external missiles. Experience and test data indicate that the burst of a ductile disc tends to generate a small number of relatively large fragments, whereas brittle discs and wheels produce a large number of pieces with a wide variety of sizes (Reference 2). To approximate the expected number of missiles

following a hypothetical nuclear turbine wheel burst, sixteen fragments in four size classes are postulated as follows: two 120-degree sectors, one 60-degree sector, three fragments weighing about one-third of the 60-degree sector, and 10 smaller pieces. It is expected that the wheels will exhibit ductile behavior and produce less than 16 fragments in the unlikely

event of bursting. However, the use of the large number can be viewed as containing an allowance for smaller missiles resulting from buckets and other parts. Postulated missile dimensions and the missile data (geometries, weights, exit energies and speed, and range of exit angles) are given in Table 3.5-6 and Table 3.5-7 (References 2 and 4).

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 11 2. Analytical Models (a) Energy Lost in Penetrating Low Pressure Turbine Casing The wheel is assumed to fracture into various fragments, with the fracture surfaces occurring in an axial-radial plane. The total kinetic energy of these fragments is reduced in collisi on with the casing structure. The methods used in determining this loss of energy are discussed in Reference 3, and are summarized below. It is conservatively assumed that there is no loss of energy during the generating of fragments, and th at there is no transfer of energy between fragments due to collision subsequent to the burst. Simple containment tests conducted by the turbine manufacturer indicate that the local shear deformation rather than gross deformation is the principal mode of energy absorption. It is assumed that the larger wheel fragments impact with and transfer energy to the stationary components of the turbine casing. An "Effective Translational Energy" is used to account for the rotational energy of the postulated missile fragments. The Stanford formula is employed in calculation of energy absorbed by the turbine casing. The wheel fragments are assumed to be oriented in the direction of minimum projected area at the instant of impact to minimize energy loss in penetration and thus, maximize the escape energy. (b) Energy of Emerging Missiles Exit energy and velocity ranges of emerging missiles are given in Table 3.5-7. Smaller fragments are visualized as either escaping through a hole in the turbine casing created by a larger fragment or

being slowed down or stopped by the stationary turbine components. The initial size and number of fragments and the missile trajectories are dependent upon the initial burst speeds and the energy adsorption capability of the stationary components.

The assumptions used in obtaining the missile exit parameters result in conservative estimates of these parameters.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 12 (c) Missile Ejection Angles Missiles generated by interior stages are ejected within a narrow angular zone (-5.0 to 5.0) with respect to the turbine axis. End stages, on the other hand, are ej ected into angles between 0.0 and 25.0 with respect to the turbine axis, away from the adjacent intermediate stages. In addition, the low pressure turbines rest upon massive reinforced concrete supports, 10 ft thick. These supports will effectively screen out all missiles emerging at angles below -14.5 (with respect to the horizontal). c. Target Description The characteristics of the safety-related areas subject to unacceptable damage by turbine missiles are provided in Table 3.5-8. Discussions of specific treatments are provided below. Note that no safety-re lated structure in the plant lies within the -5 to 5 direct missile zone for missiles from intermediate stages. 1. Containment The containment is modeled as a flat-topped cylinder having the same horizontal and vertical projected area as the domed structure. The presence of the secondary containment is ignored in the assessment of damage probabilities, since it will have little effect upon the very energetic missiles postulated above. Those turbine missiles identified as possessing sufficient energy to perforate the containment shell have been evaluated as part of a probabilistic study which has determined that these missiles fall into an

established category of accepta ble risk (see Updated FSAR Subsection 3.5.1.3d) and, as such, are not a design consideration for the containment structure. The remaining turbine missiles do not possess sufficient energy to perforate the containment shell. Some may possess sufficient energy to cause dislodgement of concrete on the inside face, local to the impact area, but the lin er plate will serve to contain these concrete fragments, thus preventing any secondary missiles from entering the containment.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 13 2. Control Building The damage criterion for the Control Building is taken to be barrier spall. No credit is taken for redundancy of vital systems in the 21.5 ft and 50.0 ft levels, since these redundant trains are not separated by missile barriers. 3. Diesel Generator Building The diesel generators are located at the 21.5 ft - level and are separated by a 2 - foot thick missile barrier. It is not possible for any direct missile to penetrate an external wall protecting one train and continue on to damage directly, or by secondary missiles, the second train. At the 51.5 ft - level, the redundant trains are not separated by missile barriers so that the damage criterion is taken to be small. 4. Condensate Storage Tank The condensate storage tank consists of an inner stainless steel tank surrounded by a reinforced concrete barrier which acts as an outer tank. This dual tank arrangement is required to contain a minimum of 200,000 gallons of water. The elevati on below which perforation of the outer tank is unacceptable is 47.5 feet (which includes margin to compensate for missile effects). It is conservatively assumed that perforation of the outer concrete tank will generate spall fragments which will perforate the steel tank. 5. Other Buildings The damage criterion for other buildings (Emergency Feedwater Pump Building, Equipment Vaults, Service Water Pumphouse, Primary Auxiliary Building) is taken to be spall of the external missile barriers. d. Turbine Missile Strike and Damage Probability Analysis Turbine missiles strike probability (P

2) is calculated using the UE&C computer code "TURB1," which is based upon the methods developed in Reference 5. For a given set of missile ejection conditions, the fraction of permissible trajectories striking any portion of a safety-related building is determined. Suitable averages over the ejection conditions are taken to obtain the estimates of P
2.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 14 The probability of target damage, given a missile strike, P 3 , is obtained by determining the fraction of missile traject ories striking a barrier which can result in damage with radiological consequences exceeding regulatory. In most cases, the damage criterion used is barrier spall and the stopping power of intermediate barriers is determined by Modified NDRC Missile Damage Equations (Reference 6). 1. Analytical Model It is assumed that one, and only one , turbine stage will fail under any set of overspeed conditions. The basis for this assumption is that the failure of the stage in question will result in an immediate end to the overspeed condition responsible for the forces resulting in stage failure. Also, since missiles are directed perpendicular to the turbine axes, it is unlikely that a missile from a failed stage will strike an adjacent stage, thus possibly resulting in cascading failures (Reference 1). The basic equation for the analysis of turbine missiles is:

P 4 = P 1 P 2 P 3 (1) where P 1 is the rate of turbine failure (events per year) resulting in the generation and ejection of missiles, P 2 is the conditional probability that a missile will strike a specified target , given its generation and ejection, and P 3 is the conditional probability that the missile will cause damage to the target that may lead to unacceptable c onsequences, given that the target is hit. P 4 is then the probability (events per year) of occurrence of unacceptable damage from a turbine failure. Two estimates of P 1, provided by the turbine manufacturer (Reference 1), are available, as shown below:

P 1 (Normal overspeed) Low Trajectory = 1.1x10

-6/year P 1 (Normal overspeed)

High Trajectory = 3.7x10

-5/year (2)P 1 (Runaway overspeed) Low Trajectory = 1.1x10

-6/year P 1 (Runaway overspeed)

High Trajectory = 3.7x10

-5/year S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 15 The estimates provided by the turbine manufacturer are based upon a detailed evaluation of the probabilities of the failure mechanisms which may lead to the various overspeed conditions. The entire failure probability is conservatively applied to each estimate. The analysis uses stage 7A only for low trajectories and all stages for high trajectories. The specific criteria for acceptability of the level of risk from turbine-generated missiles may be stated as: (a) High Trajectory Missiles The probability of a high trajectory missile strike on any safety-related area (i.e., any structure in which missile induced damage may lead to radiological consequences in excess of regulatory limits) shall be less than about 10

-7 per year per area. (b) Low Trajectory Missiles The probability that a low trajectory missile strike results in radiological consequences exceeding regulatory limits should be less than about 10

-6 per year per unit for all safety-related areas in that particular unit. This criterion is contingent upon the demonstration that the method of calculation of probabilities is

conservative. In the present analysis, the probability estimates are based upon SRP 3.5.1.3 turbine failure probabilities.

In view of the Turbine Overspeed Protection Systems described in Section 10.2, and because of the extensive quality assurance program of the manufacturer, these failure rates are considered to be most conservative. Additional conservatism is due to the assumption that barrier damage will generally lead to radiological consequences exceeding regulatory limits.

No credit is taken for redundancy, unle ss redundant trains are separated by missile barriers which would contain spall fragments.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 16 It is assumed, that the failure of a given stage will result in the emission of a number N L of k types of missiles characterized by a weight, W

i. The probability that the i-th type of miss ile will strike a given target depends upon the energy, E, and angles of em ission. The probability distribution for energy, P E, is taken to be uniform.

P(E i)dE i=dE i/(E 2i-E 2i),E 1i E i E 2i (3)= 0 ,otherwise The probability that a missile will be ejected into a solid angle element, defined by the polar angle , that (measured from the perpendicular to the turbine axis) and the azimuthal angle (measured around the polar axis) is also taken to be uniform (see Figure 3.5-2).

P'd'=cos dd/[2(sin U-sin L)],LU (4) = 0 ,otherwise The upper and lower angle constraints are dependent upon the specific emission conditions. Table 3.5-7 contains values of N L ,W i ,E li ,E 2i ,k,U , and L for each stage type for the two sets of overspeed conditions (120%

and 180%). It is necessary to relate the solid angle of emission, d', to the solid angle of emission, d, in the coordinate system in which the azimuthal angle, , is measured horizontally from the turb ine axis and the polar angle, ø, is the angle between the initial velocity vector and the horizontal plane. (See Figure 3.5-2 for angle definitions.) Thus, d'(,)=J()d(,ø) (5) where J is the Jacobian of the transformation. It is found that:

cos dd = cos ødød (6) where: cos = cos øcos (7)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 17 By combining equations (4), (5), (6) and (7), it is found that: 14.5 andU sincos cosLsin,L sinu sin 2cos os d dP (8) The additional condition in Equation (8) arises from the barrier present at the turbine. It is now necessary to relate the target coordinates to the emission angle ø. Assume that a target area increment is located at polar coordinates (, R, Y) and the turbine stag e at coordinates (O, O, Y T). Also, V i 2/g = 2 E i/W I (9) where V i is the missile ejection velocity and g the acceleration of gravity. The trajectory angle, ø, required to hit the target element is: 4 1 2 i i 2 i i i i RYTY W 4E 1RW 2ERW 2E tan (10) The positive sign in Equation (10) reference to "high trajectory" missiles, the negative sign to "low trajectory" missiles. The constraints given in Equation (6) may represent missile st rikes of either type, dependent upon the location of the area element. Equation (10) also implies a range constraint in that the interior of the square root must be nonnegative. A

ground-level target area whose coordinates are such that a low trajectory missile may strike the target is said to be in the direct zone of that missile, or, since the emission angle constraints are the same for all missiles

ejected by a given stage, the target is in the direct zone of the stage.

Targets in the direct zone of any turbine stage are considered to be in the

direct zone of the turbine unit, as shown in Figure 3.5-1. Targets outside the direct zone of the turbine may be struck by low trajectory missiles under certain conditions. Such strikes are considered direct strikes for the purposes of this program. In the fo llowing, let J equal conditions of the missile strike, and consist of the following levels: (a) Turbine number (b) Building number (c) Type of trajectory (low, high)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 18 (d) Type of hit (direct, indirect) (e) Portion of building struck (wall, roof). Let N be the number of stages (42 for 180 percent overspeed, 6 for 120 percent overspeed in the present cas e). The probability of a strike P 2 (J) may be written, EXl, Eli,Aj, dA ddA PE dEi N11 k11 N s N s 1 J P 2i i li 2 k (11) where the interior integral is performed over all building area increments such that a hit of type (c), (d) and (e) is permissible, given the constraints of Equations (8) and (10). These constraints are to be applied at the coordinates x i representing the position of the area increment with respect to the postulated failed disc 1. An expression similar to Equation (11) may be derived for P 2 (J) P 3 (J) by changing P to P' where: P' = QP (12) P 3 (J) is the probability of damage, given a missile strike, and Q is the conditional probability that a missile whose parameters permit a strike on the target increment will result in unacceptable damage, as determined by the modified NDRC Equations shown below:

G(X/d) = kNd0.2 (w i/d 3)(V i'/1000)1.8 (13) where V'i is the component of velocity normal to the surface of impact (ft/sec), d is the mean effective diameter of the missile in inches, N = .84 for blunt missiles, and

'f180/k c (14) where f c' is the concrete strength in psi. Then:

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 19 1dXG,dXG11dXG,dXG2dX (15) In the above, x is the penetration distance for the missile. Normally, d is taken to be the diameter of the circle whose area is the mean area of the missile fragment. For the missiles of interest in this program, the target damage depends only slightly upon the value of d used. Equations (13) - (15) are used to determine the effectiveness of intermediate barriers in stopping missiles by determining the amount of

energy lost in the barrier. For the purpos es of this analysis, only structures specifically designed as missile barriers are postulated to have an effect upon missile trajectories. Most intermediate vertical barriers are most likely to be struck by missiles with high angle of incidence (i.e., greater than 65 with respect to the surface normal). Although such missiles may result in unacceptable damage to those systems protected by the intermediate barrier, it is most unlikely that these missiles will cause damage after a second impact. Additional clarification of this point is provided in Subsection 3.5.1.3c. In several cases, the specific criterion used to determine acceptability of damage is that the missile barrier shall not be perforated. The modified NDRC Barrier Perforation Equations are: 1.35dX,dX0.718dX3.19d t 2 p 1.35dX,dX1.241.32 (16) The barrier thickness required to prevent perforation is t

p. The spall distance (the thickness of concrete required to prevent spall for missiles of a given type) is given by: 0.65dX,dX1.362.120.65dX,dX5.06dX7.91d t 2 s (17)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 20 The damage probability Q is assumed to be: tstp,0,tstp,1, Q (18) where t is the thickness of the barrier at the missile impact point.

Equations (13), (16) and (17) ar e applied on an area by area basis. 2. Application of the Model The computer program TURB1 consists of a set of nested loops designed to perform the integrals, averages, summations and damage and strike categorizations implied in Equation (11). Each stage/building combination is divided into the area increments by angle (approximately one degree in extent) and vertical or radial dimensions (five increments in either case). Thus, the containment for Unit 1 is divided into approximately 34 angle increments. The initial missile energy range is modified to eliminate those portions for which th e angle constraints preclude a hit on the target area in question. This modi fied range is then divided into five increments. Each combination of energy and area elements is then tested to determine which types of hits (and damage) are permissible according to the angle constraints. Sensitivity analyses conducted on the angle, lateral dimension and velocity incrementing, indicate that this increment size is sufficient to reduce computational error to a few percent. 3. Missile Shielding In this analysis, it is assumed that only reinforced concrete structures specifically designed as missile barriers can have any effect upon missile trajectories. In many cases, a safety-re lated structure having such a barrier lies on the direct line between the turbine and another safety-related building. Any low trajectory miss ile striking the second building will necessarily result in unacceptable damage to the first building. To avoid double-counting of unacceptable missile strikes, it is necessary to take

such shielding into account.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 21 The condensate storage tank is a special case, since impacts on its surface above 47.5 feet will not cause unacceptable damage. In this case, it is assumed that missiles can penetrate the condensate storage tank (losing energy in the process) and continue on to possibly damage other safety-related buildings. The total en ergy lost is taken to be the energy required to penetrate two concrete ba rriers (as calculated by the Modified NDRC Penetration Equations) each two feet thick. No credit is taken for oblique impacts or the stopping power of the inner tank. In some cases, it is possible that a missile will strike a building wall at an oblique angle and ricochet without causing any damage to the building. The building arrangement ensures that it is most unlikely that such deflections can impact on another safe ty-related structure. This point ensures conservatism in the analysis. 4. Results The results of the analysis are shown in Table 3.5-9, (low trajectory missiles) and Table 3.5-10 (high trajectory missiles). The total probability of unacceptable damage to safety-related structures from low trajectory missiles is less than 1x10

-7 per year. The probability of a high-trajectory missile hit is less than 1x10

-7 per year for any single structure. Many such missile strikes will not result in unacceptable damage. It is concluded, therefore, that the plant layout and design provides adequate protection from turbine missiles. e. Turbine Overspeed Protection A description of the Turbine Overspeed Protection System in terms of redundancy, diversity, component reliability , and testing procedures, is provided in Subsection 10.2.2. f. Turbine Valve Testing A discussion of the turbine valve tes ting procedures and test frequency is provided in Section 10.2.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 22 g. Turbine Characteristics Turbine data pertinent to the evaluation of its failure characteristics, including a description of its overall configuration, major components (e.g., steam valves, reheaters, etc.), rotor materials and their properties, steam environment (e.g.,

pressure, temperature, quality, chemistry), and other appropriate properties are provided in Section 10.2. Turbine ope rational and transient characteristics, including turbine startup and trip environment, as well as its overspeed parameters are also provided in Section 10.2. 3.5.1.4 Missiles Generated by Natural Phenomena Possible design basis tornado-generated missiles th at could originate at the site are listed in Table 3.5-11 along with their dimensions, weights and associated impact velocities. This missile spectrum is taken from Standard Review Plan 3.5.1.4, Rev. 1, Missiles Generated by Natural Phenomena (6/77), paragraph 5, designated as "SRP 3.5.1.4, Rev. 0, Missile Spectrum." (This spectrum is identical to that defined as "SRP Subsection 3.5.1.4, November 24, 1975 Missile Spectrum A" in NUREG-0800, SRP 3.5.1.4, Rev. 2, July 1981.) These missiles are considered to be capable of striking in all directions with vertical velocities equal to 80 percent of the acceptable horizontal velocities. Missiles A, B, C, D and E are to be considered at all elevations, and missiles F and G at elevations up to 30 feet above all grade levels within one-half mile of the facility structures.

The structures and/or barriers designed to resist tornado-generated missiles are tabulated in Table 3.5-12. Information on the protected systems or components is included in Subsection 3.5.2. All walls and slabs used for missile protection are constructed of reinforced concrete having a minimum thickness of two feet and a minimum specified 28-day compression strength of 3000 PSI. The containment enclosure structure, which is constructed of reinforced concrete with a minimum thickness of 15 inches, will not be breached by tornado-generated missiles.

The design procedures discussed above will ensure that the probability of a tornado missile impact on a missile barrier leading to exceeding 10 CFR 100 limitations will be less than 10

-7 per year. In addition, an evaluation (Reference 20) of a missile entering a safety-related structure resulted in a probability of about 10

-6 per year.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 23 3.5.1.5 Missiles Generated by Events near the Site As stated in Subsection 2.2.3.1, missiles generated by explosions near the site are extremely unlikely to reach the site area. Therefore, no missile impacts from these sources need be postulated. 3.5.1.6 Aircraft Hazards

a. The Model As discussed in Subsection 2.2.2.5, it has been determined that the method used by the NRC (Reference 8) in its evaluation of the Boardman Nuclear Plant is applicable to the Seabrook site. The model treats each aircraft operating from an airport or on an airway individually, finds the probability of impact for each plane and sums up all the individual probabilities to obtain the total crash probability:

i Transit Random Exposure R i A i K1i N i P (1) where i = Summing index for K different aircraft N i = Number of flights of aircraft "i" per year that have a finite probability of impacting the site A i = Area of critical structures at the site (mi

2) R i = Accident rate of aircraft "i" (acc/hr) i Transit Random Exposure Time per square mile that aircraft "i" is in the vicinity of the site (hr/mi
2)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 24 b. Area of Critical Structures (A)

The area is determined by assuming a thirty degree slope for a disabled aircraft. The structures considered as a target include:

1. Containment Building 2. Fuel Storage Building
3. Primary Auxiliary Building
4. Control Building
5. Diesel Generator Building
6. Refueling water storage tank
7. Reactor makeup water storage tank The Turbine, Heater Bay, Waste Processing and Administration Buildings are not considered to be targets and are assumed to serve as a shield for critical buildings behind them. Figure 3.5-3 is a sketch of the plot plan and shows the effective target area. The target area shown on this figure was reduced by eliminating the Service Water Pumphouse which is backed up by the ultimate heat sink cooling tower 1000 feet away. Eith er one of these two structures can supply cooling water. Therefore, because of this redundancy and separation, they were not considered in the target area. The resultant target area is 0.005 square miles per unit. The Containment Building is considered as part of the target area for aircraft weighing more than 81,800 pounds. Base d on a structural analysis, the containment can withstand the impact of an FB-111A and smaller aircraft.

Therefore, the target area has been redu ced for these aircraft to an area of 0.0029 square miles to include the other cr itical structures only. The structural analysis, presented in Appendix 2P, verifies the containment integrity after an 81,800 pound aircraft impact of the FB-111A type. It was also shown by analysis that all the critical structures can withstand the impact of small aircraft equal to or less than 12,500 pounds.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 25 c. Exposure Per Random Transit The exposure per random transit is vgh 2 1 , where "v" is the speed of the aircraft while in the vicinity of the site, "g" is the glide ratio of the particular aircraft and "h" is the altitude of the aircraft. A glide ratio of 17 has been assumed for all military and commercial aircraft excepting the FB-111A. A glide ratio of 13 is assumed for the FB-111A. d. Federal Airways A description of the federal airways and a conservative estimate of the number of aircraft that pass within 2 miles of Seabrook can be found in Subsection 2.2.2.5. FAA area specialists (Reference 9) examined routing slips of each aircraft that flew in the Pease Sector on the peak traffic day of 1979 (July 27, 1979). They determined that 224 flights could have actually passed within 2 miles of the site.

Of the 224 flights, 128 aircraft weighed more than 12,500 pounds. A cruise airspeed for each aircraft was determined (Reference 10), and from this an average airspeed was computed. The average altitude for the 128 aircraft was 30,000 feet. Based upon 128 possible overflights of aircraft over 12,500 pounds on the peak traffic day, an annual estimate of 46,720 overflights can be computed. For commercial aircraft:

Exp/Trans vgh 2 1 6818.5175002 1 mihr101.04 2 5 e. Airports Within 5 Miles Hampton Airport is described in Subsection 2.2.2.5. As stated, the aircraft at Hampton are single engine and theref ore weigh considerably less than 12,500 pounds. It can be assumed that since the airport consists of a very short turf runway, it is highly unlikely that an aircraft weighing more than 12,500 pounds would use this airport. Sin ce the critical structures at Seabrook can withstand the impact of aircraft weighing less than 12,500 pounds, the risk from this source is zero.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 26 f. Airports Beyond 5 Miles Subsection 2.2.2.5 describes airports beyond 5 miles and their associated risk to the Seabrook site. g. Military Installations A description of the traffic patterns, aircraft, accident rates, and annual number of transits proximal to Seabrook for each type aircraft using Pease Air Force Base can be found in Subsection 2.2.2.5. The traffic pattern altitude which brings aircraft operating out of PAFB in proximity to Seabrook is 2,000 feet. A downwind airspeed of 220 miles per hour is assumed for the FB-111A, and 200 miles per hour for all other military aircraft. For FB-111A aircraft:

Exp/Trans vgh 2 1 379.132202 1 mihr104.61 2 4 For all other military aircraft:

Exp/Trans 379.172002 1 mihr103.89 2 4 h. Accident Probability Using equation (1), the total accident probability is calculated to be 9.71x10

-8. Table 3.5-13 shows the inputs to the calculation of this result. The number of operations and accident rates for each type of aircraft can be found in Subsection 2.2.2.5.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 27 3.5.2 Structures, Systems and Components to be Protected from Externally Generated Missiles All plant structures, systems and compone nts whose failure could lead to offsite radiological consequences, or which are required to shut down the reactor and maintain it in a safe condition while assuming an additional single failure, are designated ANS Safety Class 1, 2 or 3 and/or seismic Category I, and are listed in Table 3.2-1 and Table 3.2-2. Safety-related components, including essential piping, instrumentation and electrical equipment, are protected against damage from externally generated missiles by physical barriers or protective structures. The structures protecting the systems and components important to safety are listed in Table 3.5-12; plan and elevation drawings pertinent to these structures are found in Section 1.2. Discussions on design requirements which exempt the refueling water storage tank and spray additive tank from mi ssile protection are presented in Subsection 6.2.2. Protection of the Fuel Storage Building spent fuel pool cooling/cleanup system and storage pool from missiles which could penetrate the nonmissile-proof fuel shipping cask rail car access door is provided by an interior missile-proof wall, as shown in Figure 1.2-16. Results of the probabilistic analysis (Reference 20) indicate that the overall entrance-probability for the Fuel Storage Building is less than 3x10-9 per year. The ultimate heat sink complex (Subsection 9.2.5), which consists of the mechanical draft cooling towers and the Atlantic Ocean service water system, has appropriate portions protected against all credible missiles. Those portions of the

intake and discharge transition structur es which house the safety-related service water valves (valve pit area) are designed to withstand tornado generated missiles.

Since the safety-related portions of the piping entering the transition structures

are enclosed and protected as they pass through this area, only the pipe entrance to the structure is exposed. However, the pipes are located under water at or

below elevation (-)35'-0". The pipes are, therefore, not exposed to credible tornado-generated missiles.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 28 Referring to Standard Review Plan 3.

5.1.4, "Missiles Generated by Natural Phenomena," the largest missiles which might be considered to potentially block the pipe entrance are the 35-foot long utility pole and the 4000 lb. automobile. In order to block the pipe entrance, the automobile would have to be directed through the top of the transition structure, missing the concrete cross-bracing, and

drop exactly at the pipe location, sinking to the elevation of the pipe. Since the pipe is located at elevation (-)40'-0" (in the intake transition structure), and the floor is at elevation (-)55'-0" the automobile would continue to sink to the floor where it could not pose any blockage to the pipe. The utility pole would have to be similarly directed, however, it would pose no blockage problem since it would float at water level considerably above the elevation of the pipe. The only safety feature of the cooling tower is to function in the event of a seismic disturbance of sufficient magnitude to interrupt the flow in the main circulating water tunnels. Accordingly, only those parts of the Service Water Cooling Tower structure which protect service water piping up to and including the cooling tower pump discharge valves require missile protection. Certain parts of the tower and its structure are not protected against missiles. These include the tower air intakes, tower pump, fans and gear boxes and electrical switchgear. The remainder of the service water equipment and piping is located within the missile-protected Service Water Pumphouse, the Primary Auxiliary Building, or buried underground. The sink safety func tion is assured through the use of the missile-protected Service Water System and its seismic Category I cooling tower

which provides an alternate source of cooling water. Refer to Subsection 9.2.5 for further details concerning the ultimate heat sink. 3.5.3 Barrier Design Procedures The structures and barriers identified in Subsection 3.5.2 are designed to withstand the local effects and overall effects of the applicable missiles. The following areas are

discussed: a. Methods for the prediction of local damage in the impacted area, including estimation of the depth of penetration and, in case of concrete barriers, the potential for generation of secondary missiles by spalling or scabbing effects. b. Methods for the prediction of the overall response of the structure or barrier due to the missile impact. This includes assumptions on acceptable ductility ratios where elasto-plastic behavior is relied upon, and procedures for estimation of forces, moments, and shears induced in the barrier by the impact force of the missile.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 29 3.5.3.1 Local Damage Prediction

a. In Concrete The modified Petry equation, as given by A. Amirikian (Reference 11), has been commonly used to estimate missile penetration into concrete barriers. Sufficient thickness of concrete is required to prevent perforation and scabbing. Modified Petry's formula for penetration is:

D = KA p V' where: D = penetration in a barrier of thickness, t, where t 3D, both t and D in feet K = coefficient for penetration for the material = 3.50x10

-3 for reinforced concrete (f' c = 3000 psi)

A p = sectional pressure = W/A lbs./sq ft W = weight of missile in lbs.

A = cross section of missile in ft 2 V' = 215,000 V 1 log 2 10 V = velocity of missile in ft/sec For barriers having thickness t < 3D, the penetration D' was calculated as: 2a'4e1DD' where: a' = t/D The minimum thickness required to prevent scabbing for steel missiles having solid circular cross section was determined from Figure 3.5-4. For other noncompressible missiles, design for no scabbing was based on t 3D, Reference 11.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 30 Since the PSAR, experimental data ha s been generated by Sandia Laboratory for the Electric Power Research Institute for the specific missiles, velocities and targets applicable to these structures and barriers. The semi-analytical, semi-empirical modified National Defense Research Committee (NDRC) equations have been found to be more accurate than the modified Petry equations, Reference 12. The Modified NDRC formula for penetration is:

1.8 1000d

VDWNK4x 2.0dxfor where: x = penetration depth in inches K = c f'180/where f' c is the concrete design strength N = missile shape factor = 0.72 for a flat nose missile W = weight of missile in lbs.

d = diameter of missile in inches V = striking velocity of missile in ft/sec The minimum thickness to prevent scabbing is given by:

t s/d = 2.12 + 1.36 x/d for 0.65 x/d 11.75 where for pipes: d = d e = effective diameter as though a solid cylinder S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 31 A factor of safety of 1.2 was applied to t

s. The design thickness of walls and roof slabs was checked for its adequacy to prevent scabbing using both sets of equations and was found to be satisfactory. b. In Steel Tornado doors are 2" thick A-36 st eel barriers designed to resist tornado-generated missiles in addition to tornado wind pressures (Reference 21). Maximum penetration depths are calculated by the BRC formula (Reference 15) to be 0.84 inch for the 6" diameter steel pipe and ~0.80 inch for both the 12" diameter steel pipe and 1" diameter steel rod. 3.5.3.2 Overall Damage Prediction Concrete barriers and structures subject to missile impact were designed to sustain the dynamic force of impact without shear or flexural failures, or excessive deformation. The dynamic force is a function of many parameters including the mass, velocity and rigidity of the missile, the natural period of vibration, stiffness and ductility of the target. The missiles can be conveniently grouped into two basic categories, rigid (nondeforming) and nonrigid (deforming). Examples of each group include turbine parts and tornado-borne pi pes and rebar for the first, and aircraft and tornado-borne wood poles and automobiles for the second. A structure or barrier impacted by a missile of either group was analyzed for its maximum response to a force-time loading function using the methods in References 13, 14 and 15. The problem is idealized as an equivalent one-degree-of-freedom elasto-plastic structure subjected to a suddenly applied load of given pulse shape, duration and magnitude. The force-time load histories for the rigid pipe and rebar missiles are calculated using a modified Williamson and Alvey method (References 15 and 16),

while those for the wood pole and aircraft are calculated per Riera (Reference 17), and that for the auto per Bechtel's BC-TOP-9A (Reference 18). a. Missile Force-Time Histories

1. Pipes, Rebar and Other Rigid Missiles that Penetrate The missile is assumed to penetrat e the barrier with linearly decreasing velocity and constant force. Th e resulting rectangular pulse forcing function has a magnitude, F, and duration, t d , calculated as:

x V g w 2 1 F 2 o S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 32 and V 2 t o x d where: W = weight of missile (lbs)

V O = impact velocity (ips) g = 386.4 in/sec 2 x = penetration depth (in)

The penetration depth was calculated by the modified NDRC equation which has been shown to give the best correlation to experimental data. The pulse magnitudes and durations for the 743 lb., 12" steel pipe are as follows: Direction Velocity (fps) F (Kips) t d (sec.) Horizontal 211 743 0.00656 Vertical 169 582 0.00671

2. Utility Pole A force-time history was developed for the utility pole using the Riera formula, Reference 17 as follows: F(t) = F c A + V 2 (t) where:

F c = the "crushing strength" of the wood pole (lb./ft

2) A = the cross-sectional area (ft
2) = the mass per unit length (slug/ft)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 33 V(t) = the instantaneous velocity of the missile during impact (ft/sec) Experimental data from the Sandia tests for EPRI, Reference 19, were used to define a reasonable value of crushi ng strength. The above equation was integrated in time for the definition of F(t). The forcing function was then approximated by a rectangular pulse as:

Direction Velocity (fps) F (Kips) t d (sec.) Horizontal 211 202.5 0.054 Vertical 169 192 0.044

3. Auto The force-time history defined in Bechtel's report BC-TOP-9A, Reference 18, was used for the auto impact. F(t) = 0.625 V s W m sin 20t 0 t 0.0785 sec = 0 t > 0.0785 sec where:

V s = the initial impact velocity (ft/sec)

W m = weight of auto (lbs.) This is a quarter sine wave with peak force of 265 Kips and duration of 0.0785 seconds for V s equal to 106 ft/sec. 4. Aircraft For a discussion on aircraft hazards and the ability of seismic Category I structures, particularly the containment, to withstand postulated impacts from aircraft, see Subsection 3.5.1.6, Aircraft Hazards. b. Procedure for Calculating Overall Response The procedures used in calculating struct ural response of walls and slabs to the impact forces of the rigid missiles, u tility pole and auto are those described in References 13, 14 and 15, summarized as follows:

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 34 1. The load pulse shape, peak force, F, and duration, t d , are defined as above. 2. The slab (or beam) capacity, R m , natural period of vibration, T, and maximum allowable ductility, , are calculated or defined. 3. The effect of other forces acting simultaneously with the impact force on the structure or barrier are included. The displacement, X s , of the structure under these static loads and an effective point force, R s , are calculated. 4. The capacity of the structure to resist the impactive force is defined as R' m = R m-R s and an effective allowable ductility is defined as '= (X m-X s)/(X e-X s) where X e is the yield displacement and X m = X e. 5. The system behavior is assumed elasto-plastic with an equivalent single degree-of-freedom. Ratios C T = t d/T and C R = R'm/F are formed. Figure 3.5-5, a chart of X m/X e curves for rectangular impulse loads, or similar figures for the appropriate pulse shape are entered with the values of C T and C R, and the displacement response measure, X m/X e , is determined. This measure is compared to as follows:

X m/X e  : no failure, acceptable > : unacceptable Additionally X m/X e 1 : elastic response >1 : permanent displacement The relative structural response to the tornado-borne rigid missiles, pipes and rebar, were compared. The 12" pipe produced the greatest deformations in all cases. Hence, subs equent analysis of overall structural effects disregarded the smaller pipes and rebar.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 35 c. Permissible Ductility Ratios Reference 14 was followed for definition of permissible ductility ratios of reinforced concrete members. 1. For beams, walls or slabs where flexure controls design:

= 0.05/( -') 10 where: = A s/bd, ' = A s'/bd A s = area of tension reinforcement A s' = area of compression reinforcement d = distance from extreme compression fiber to centroid of tension reinforcing steel b = width of beam or unit width of slab For flexure to control design, the load capacity of a structural element in shear shall be at least 20 percent greater than the load capacity in flexure.

Punching shear capacity of reinforced slabs under dynamic loading has been shown to be greater than that under static loading. The ultimate shear stress was assumed as V u = 4.4 cf'Rigid missiles (e.g., pipes and rebar) which are shown not to perforate a barrier or structure shall be considered as producing flexural controlled response in the structure. 2. For beams, walls and slabs where shear controls design and shear is carried by concrete alone, = 1.3. 3. For axial compression, = 1.3.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 36 The following permissible ductility ratios were used for steel members: 1. For flexural members, = 0.2 u/y where:

u = ultimate strain at rupture y = yield strain 2. For tension members, = 0.5 u/y 3. For compression members, = 1.0 d. Structural Capacity and Period of Vibration The ultimate moment capacity of walls, slabs and beams is calculated using provisions of the appropriate codes.

The collapse load for a point load was related to the ultimate moments by yield-line theory as in Reference 15. The natural period of vibration of the "equivalent" single degree-of-freedom structure was calculated to correspond to that of the first flexural mode of the actual structure. The effective moments of inertia of concrete beams, walls and slabs were assumed to be the average of the gross and cracked moments of inertia.

Tornado doors are steel barriers designed to resist tornado-generated missiles in combination with tornado wind pressures (Reference 21). Energy and momentum principles (Reference 15) were used to determine the maximum equivalent static loads due to missile impact. The 12" diameter steel pipe missile was found to produce the worst case load which was subsequently used in all analyses. Linear elastic finite element analyses of the total door system, including hinges, determined deflections, forces a nd stresses up to first yiel d in the door. A yield line analysis was performed to assess maximum capacity allowing for elasto-plastic behavior.

Ductility up to 20 was considered acceptable (Reference 15). The hinges and latches were evaluated for maximum shear and axial forces.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 37 3.5.4 References

1. Missile Probability for Nuclear High Pressure ADSP With Thermal Uprate, Florida Power and Light Company, Seabrook #1, 170X683 in 170X560 LPA, LPB, and LPC Rotors, June 2004. 2. J. E. Downs, "Hypothetical Turbine Mi ssile Data, 43-inch Last Stage Bucket Units," General Electric Company Memo Report, March 15, 1973. 3. D. C. Gonyea, "An Analysis of the Energy of Hypothetical Wheel Missiles Escaping from Turbine Casings," Technical Information Series No. DF73LS12, General Electric Company, February 1973. 4. J. E. Downs, "Hypothetical Turbine Missiles - Sample Calculations," General Electric Company Memo Report, March 17, 1973. 5. A. K. Bhattacharyya and S. K. Chaudhur i, "The Probability of a Turbine Missile Hitting a Particular Region of a Nuclear Power Plant," Nuclear Technology, Vol.

28, February 1976. 6. R. P. Kennedy, "A Review of Pro cedures for the Analyses and Design of Concrete Structures to Resist Missile Impact Effects," Journal of Nuclear Engineering and Design, Vol. 37, No. 2, May 1976. 7. S. H. Bush, "Probability of Damage to Nuclear Components Due to Turbine Failures," Nuclear Safety Vol. 14, No. 3, May-June 1973. 8. "Aircraft Consideration Preapplicati ons Site Review by the Directorate of Licensing, U.S. Atomic Energy Commission," in the matter of Portland General Electric Company, Boardman Nuclear Plant, Boardman, Oregon, Project Numbers 4 and 5. 9. Personal Communications, Mr. Russ Scarberry, Federal Aviation Administration, Nashua, New Hampshire, January 15, 1980. 10. "Jane's All The World's Aircraft," Edited by John W.R. Taylor, London, England, 1978-79. 11. A. Amirikian, "Design of Protective St ructures," Bureau of Yards and Docks, Publication No. NAVDOCKS P-51, Department of the Navy, Washington, D.C. (August 1950).

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Missile Protection Revision 10 Section 3.5 Page 38 12. G.E. Sliter, "Assessment of Empirical Concrete Impact Formula," Paper presented at ASCE Conference on Civil Engineeri ng and Nuclear Power, Boston, Mass., April 2-3, 1979. 13. J.J. Biggs, Introduction to Structural Dynamics , McGraw-Hill, 1964, Chapter 2. 14. ACI Code 349, Requirement for Nuclear Safety Related Concrete Structure, Appendix C - Special Provision for Impulsive and Impactive Effects. 15. ASCE Structural Analysis and Design of Nuclear Plant Facilities , Manual No. 5, 1980. 16. R.A. Williamson and R.R. Alvey, "Impact Effects of Fragments Striking Structural Elements," Holmes and Narver, 1973. 17. J. D. Riera, "On the Stress Analysis of Structures Subjected to Aircraft Impact Forces," Nuclear Engineering and Design, Volume 8, 1968. 18. "Design of Structures for Missile Im pact," Topical Report BC-TOP-9A, Bechtel Power Corp., 1974. 19. "Full-Scale Tornado-Missile Impact Te sts," EPRI NP-440, Final Report, July 1977. 20. "Seabrook Nuclear Power Plant - Tornado Missile Analysis," Applied Research Associates, Inc., September 1983. 21. "Design Analysis of Double Door for Seismic and Tornado Loads," Protective Materials Co., Inc., December 1983.

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Page 1 3.6(B) PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING

a. Introduction General Design Criterion 4 of Appendix A to 10 CFR 50 requires that structures, systems and components important to safety be protected against the dynamic effects of piping failures. This sec tion discusses the design bases and design measures used to ensure that all essential structures, systems and components

located inside and outside the reactor containment, including the components of the reactor coolant pressure boundary, have been adequately protected against the

effects of possible blowdown jet and r eactive forces and pipe whips resulting from postulated rupture of piping located both inside and outside of containment. The required information is furnished in two sections: 1. Subsections 3.6(B).1 and 3.6(B).2 and respective subsections address all piping systems inside and outside containment, exclusive of the reactor

coolant loop piping. 2. Section 3.6(N) and subsections, whic h have been furnished by the NSSS supplier, address only the reactor c oolant loop piping inside the reactor containment and the loops' support system. The criteria used to identify high ener gy piping and assure the protection of essential equipment from the potential failure of nearby piping systems are per the NRC's Auxiliary Systems Branch (ASB) Technical Position ASB 3-1. The criteria employed to define and locate postulated piping breaks/cracks and their associated effects are the NRC's M echanical Engineering Branch (MEB) Technical Position MEB 3-1. The MEB 3-1 break postulation criteria supersede the more conservative criteria initially used for piping systems outside containment (ASB 3-1) and inside containment (Regulatory Guide 1.46). For the reactor coolant loop, Westinghouse Topical Report, WCAP-8082, "Pipe Break for the LOCA Analysis of the Westinghouse Primary Coolant Loop," provides the

original criteria for postulating breaks in the RC loop. The basis for eliminating eight of these postulated pipe breaks in the RC loop is provided in Reference 3. The Seabrook Station coolant system pi ping is consistent with the design considered in WCAP-8082, which has been approved by the NRC staff.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 2 b. Definitions High Energy Fluid Systems or Lines - Fluid systems or lines which, during normal plant conditions, are either in operation or maintained pressurized under conditions where either or both of the following conditions are met: - Maximum operating temperature exceeds 200 F - Maximum operating pressure exceeds 275 psig.

Moderate Energy Fluid Systems or Lines - Fluid system or lines which during normal plant conditions, are either in operation or maintained pressurized above atmospheric pressure under conditions where both the following conditions are met: - Maximum operating temperature is 200 F or less - Maximum operating pressure is 275 psig or less.

Normal Plant Conditions - Plant operating conditions during reactor startup, operating at power, hot standby or reactor cooldown to cold shutdown conditions.

Upset Plant Conditions - Plant operating conditions during system transient conditions that may occur with moderate frequency during plant service life and are anticipated operational occurrences, but not during system testing.

Essential Systems and Components - Systems and components required to shutdown the reactor and mitigate the consequences of a postulated piping failure

without offsite power.

Postulated Piping Failure - Longitudinal and circumferential breaks in high energy fluid system piping and through-wall leakage cracks in moderate energy fluid system piping postulated according to the provisions of Subsection 3.6(B).2 below. S A - Allowable stress range for thermal expansion as defined in subarticle NC3600 of the ASME Code,Section III, 1971 Edition, with Addenda up to and including Winter 1972.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 3 S h - Allowable stress at maximum temperature.

S m - Design stress intensity defined in subarticle NB-3600 of the ASME Code.

Single Active Component Failure - Malfunctions or loss of function of a component of an electrical or fluid system. A failure of an active component of a fluid system is not considered to include loss of component structural integrity. The direct consequences of a single active component failure are considered to be

part of the single failure.

Terminal Ends - Extremities of piping runs that connect to structures, components (vessels, pumps, valves) or pipe anchors that act as rigid constraints to thermal expansion. A branch connection to a main piping run is a terminal end of the

branch run. Intersections of runs of comparable size and fixity need not be considered terminal ends when so justified by the analysis. Terminal ends, for the purpose of

postulating breaks, are selected at points located immediately outside or beyond the required pipe whip restraints located inside and outside containment at

penetration areas. For lines that are pre ssurized by virtue of a connection into a pressurized, normally operating high energy line, the connected line is designated as high energy back to the first normally closed automatic or manual valve or to the first in-line normally closed check valv

e. In a piping run of this type, the terminal end is considered to be at the piping connection to the closed valve. In the event of valve leakage during normal plant operation, portions beyond this valve might become pressurized. A pipe br eak in these portions is not considered since it would not produce any sustained pipe whipping or jet impingement due to a lack of an energy reservoir.

Five Degree Restraint - A device which restrains the pipe in such a way that only axial loads can be transmitted past the restraint.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 4 3.6(B).1 Postulated Piping Failures in Fluid Systems 3.6(B).1.1 Design Bases

a. Equipment Potentially Susceptible to Effects of Piping Failure Systems and components important to pl ant safety or shutdown (referred to as essential systems and components), located proximate to high or moderate energy piping systems, which are potentially sus ceptible to the consequences of piping systems breaks and cracks, are listed in Table 3.6(B)-1. The identification of this equipment is related to predetermined piping failure locations, determined by the methodology discussed in Subsection 3.6(B).2. Table 3.9(B)-27 and Table 3.9(N)-11 tabul ate all active valves whose function must be unimpaired in the event of a piping failure. Figure 3.6(B)-1, Figure 3.6(B)-2 and Figure 3.6(B)-5, Figure 3.6(B)-6, Figure 3.6(B)-7, Figure 3.6(B)-8, Figure 3.6(B)-9, Figure 3.6(B)-10, Figure 3.6(B)-11, Figure 3.6(B)-12, Figure 3.6(B)-13, Figure 3.6(B)-14, Figure 3.6(B)-15, Figure 3.6(B)-16, Figure 3.6(B)-17, Figure 3.6(B)-18, Figure 3.6(B)-19, Figure 3.6(B)-20, Figure 3.6(B)-21, Figure 3.6(B)-22, Figure 3.6(B)-23, Figure 3.6(B)-24, Figure 3.6(B)-25, Figure 3.6(B)-26, Figure 3.6(B)-27, Figure 3.6(B)-28, Figure 3.6(B)-29, Figure 3.6(B)-30, Figure 3.6(B)-31, Figure 3.6(B)-32 show the locations of pipe whip restraints with respect to the postulated pipe ruptures and relative locations of potentially affected essential components. The limiting acceptable conditions for, and the measures taken to protect the essential systems and components, are di scussed in the pipe rupture analysis summary in Appendix 3A.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 5 b. Design Criteria for Protection Against Piping Failures The following criteria were used as guide lines during the station design to assure the protection of essential equipment from potential failure of nearby piping systems: 1. Piping Systems Containing High Energy Fluids (a) Piping systems are isolated by adequate physical separation, and remotely located from essential systems and components required to shut down the reactor safely and maintain the station in a cold

shutdown condition. (b) Where isolation by remote location is considered impractical, piping systems, or portions of the systems, are enclosed within

structures suitably designed to protect adjoining essential systems and components from postulated piping failures within the

enclosure. (c) Where both isolation by remote location and enclosure in protective structures are considered impractical, the piping systems or portions of the systems are provided with restraints and protective measures so that the operability and integrity of the structures, safety systems and components would not be impaired. (d) Protective enclosures for the piping systems are designed as seismic Category I structures capable of withstanding the combined effects of a postulated pipe break, the dynamic effects of pipe whipping, the jet impingement forces and the compartment pressurization resulting from discharging fluids in combination with the specified seismic event of the Safe Shutdown Earthquake and normal operating load.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 6 (e) Piping systems containing high en ergy fluids are designed so that the effects of a single postulated pipe break will not initiate unacceptable failures of other pipes or components. In addition, any systems, or portions of systems, that are designed to mitigate the consequences of a postulated pipe failure and place the reactor in the cold shutdown condition, ar e provided with design features that will ensure the performance of their safety function, assuming a single active component failure. (f) For a postulated pipe failure, an escape of steam, water and heat from structures enclosing the pi ping shall not preclude: (1) the accessibility to surrounding areas important to the safe control of reactor operations, (2) the habitability of the control room, (3) the ability of instrumentation, electric power supplies, and components and controls to initiate, actuate and complete a safety action. In this regard, a loss of redundancy is considered permissible, but not

the loss of function. (g) The design measures used for the protection of structures, systems, and components important to safety will not prevent in-service examinations of ASME Class 2 and 3 pressure-retaining components, as required by the rules of ASME B&PV Code,Section XI, "In-Service Inspection of Nuclear Power Plant Components." 2. Piping Systems Containing Moderate-Energy Fluids (a) Piping systems containing modera te-energy fluids are designed to comply with the criteria app lied to high-energy fluid piping systems, as stated above, except that the piping is postulated to develop a limited-size through-wall l eakage crack instead of a pipe break. (b) For each postulated leakage condition, design measures are provided that will provide protection from the effects of the resulting water spray and flooding.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 7 3. Exceptions Measures for protection against pipe whipping or jet impingement resulting from the breaks postulated in Subsection 3.6(B).2 are not provided for piping where any of the following applies: (a) Piping is physically separated or isolated from any essential system or component necessary for plant safety or shutdown by means of barriers, or is restrained from whipping by plant design features such as encasement. (b) The broken pipe cannot cause unacceptable damage to any essential system or component. (c) The energy associated with the whipping pipe can be demonstrated to be insufficient to impair to an unacceptable level the safety function of an essential system or component. For example, a whipping pipe is considered unable to rupture an impacted pipe of equal or larger nominal pipe size and equal or heavier wall

thickness.

3.6(B).1.2 Description High energy lines located in structures housing components essential for safe plant shutdown are listed in Table 3.6(B)-2. Relative to possible dynamic effects of pipe fa ilure in the Seabrook plant layout, essential systems and components are protected from the dynamic effects of rupture of high energy piping primarily by separation and redundancy. Routing of high energy lines has been arranged to provide the maximum amount of protection by using plant structural elements, such as wall or columns, and routing the high energy lines as far as practicable from essential components. In cases where separation is not possible, pipe whip restraints are used to prevent uncontrolled whipping of the high energy piping. Compartments of primary interest are the containment structure, the main steam and feedwater pipe tunnels, and the Containment Enclosure Building and its attached compartments.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 8 In the case of the control room, there are no hi gh energy lines in the area which could affect habitability as a result of pipe whip. The main steam and feedwater lines on the pipe bridge are separated from the control room by the seismic Category I Control Building wall, which has been reinforced to protect the control room environment from postulated breaks in, or whip loads from, the main steam and feedwater lines. Control room habitability systems are discussed in

Section 6.4. The high energy lines outside containment whose br eaks or cracks could have the greatest effect on environment within the structures housing components essential for safe plant shutdown are listed below: a. Primary Auxiliary Building Steam generator blowdown lines Auxiliary steam and condensate lines Chemical and volume control system letdown line Hot water heating lines b. Fuel Storage Building Hot water heating lines c. Containment Enclosure and Connected Buildings Hot water heating lines d. Main Steam and Feedwater Pipe Chase Main steam lines Feedwater lines e. Diesel Generator Building Hot water heating line f. Control Building Hot water heating line S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 9 g. Emergency Feedwater Pumphouse Hot water heating line h. Service Water Pumphouse Hot water heating lines 3.6(B).1.3 Safety Evaluation Appendix 3A summarizes the results, including environmental, of the failure modes and effects analysis of breaks/cracks in high and moderate energy piping systems in each of the structures housing essential components. Appendix 3A verifi es that the consequences of failures of high and moderate energy lines will not affect the ability of the plant to be shut down safely. The analysis considered the effects of single active component failures occurring in required systems concurrent with the postulated event. The potential effects of internally and externally generated missiles are discussed in Section 3.5.

Pressure rise analyses of structures and compartments due to piping breaks are discussed in Section 6.2 and Appendix 3I.

3.6(B).2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping This section describes the design bases for locating postulated breaks and cracks in piping situated both inside and outside of containment, the procedures used to define the jet thrust reaction at the break or crack location, and the jet impingement loading on adjacent safety-related structures, equipment, systems and components.

3.6(B).2.1 Criteria Used to Define Break and Crack Location and Configuration The criteria employed for defining break and crack locations and configurations in primary loop piping inside containment is discussed in Subsection 3.6(N).2.1. This section discusses all other plant piping.

The criteria are provided for those high and moderate energy piping systems for which separation or enclosure cannot be achieved.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 10 a. High Energy Piping

1. ASME Section III Code Class 1 Piping Breaks were postulated to occur at the following locations in each piping run or branch run: (a) Terminal ends (b) At all intermediate locations where the maximum stress range, derived on an elastically calculated basis under the loadings associated with operating plant conditions and an operating basis

earthquake, as calculated by Equation 0 and either Equations 12 or 13 of ASME Code NB-3650 exceeds 2.4S m , where S m is the allowable design stress intensity value. (c) At intermediate locations where the cumulative usage factor exceeds 0.1.

2. ME Section III Code Class 2 and 3 Piping Breaks were postulated to occur at the following locations for each piping run or branch run which does not penetrate the containment: (a) Terminal ends (b) Any intermediate locations between terminal ends where stresses as calculated by Equations (9) a nd (10) of NC/ND-3652 derived by elastic methods under the loadings associated with operational plant conditions and an opera ting basis earthquake exceed 0.8 (1.2 S h + S A). Initial break postulations using the conservative ASB 3-1 (outside containment) and R.G. 1.46 (inside containment) criteria of 0.8 (S h + S A) are conservative.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 11 See below for piping penetrating the containment.

3. Nonnuclear Piping Breaks in nonnuclear piping were postulated at the following locations in each piping run or branch run: (a) Terminal ends (b) Each structural discontinuity (elbows, tees, reducers, Valves).
4. Piping Penetrating Containment All piping penetrating the containment is ASME Section III, Code Class 2. All high energy, high temperature lines penetrating containment make use of integrally forged flued heads. A detailed discussion of the design of these flued heads is given in Reference 1. For main steam and feedwater piping penetrating containment, no breaks were postulated between the first whip restraint inside the containment and the five-degree restraint outside containment, since the following conditions are met: (a) The maximum stresses, as calculated by the sum of equations (9) and (10) in paragraph NC-3652 of Section III of the Code (for piping design, the applicable Code edition is the 1971 Code, with addenda up to and including Winter 1972), considering normal and

upset conditions and an OBE event, do not exceed 0.8(1.2 S h + S a). (b) The maximum stress, as calculated by equation 9 of paragraph NC-3652 under the loadings resulting from a postulated piping failure of fluid system piping beyond these portions of the piping, does not exceed 1.8 S

h. (c) The number of circumferential and longitudinal weld points in piping have been minimized. (d) The length of those portions of piping have been reduced to the minimum practical.

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Page 12 (e) The design of pipe restraints and anchors have not generally required welding directly to the outer surface of the pipe. Where anchors or restraints were neede d, forgings were used to avoid welding to the surface of the pipe. Where lugs were used for riser clamps, a detailed analysis was made to assure compliance with stress limits stated above. To assure the protection of essential equipment from the jet impingement and environmental e ffects of high energy line breaks, longitudinal breaks were postulated (but not concurrently) in the main steam and feedwater lines in the pipe tunnels outside containment. Each postulated longitudinal break was considered

to have a cross-sectional area of one-square foot, and was assumed to occur at a location that has the greatest effect on essential equipment. Since the primary reason for postulating the one-square foot break is for the protection of essential equipment, jet impingement on nearby structures was not evaluated. Environmental effects (pressure, temperature) were evaluated for pipe tunnel structures. Lug attachments welded to Class 2 and 3 pipes are qualified by a procedure whose equivalent methodology is more conservative than that presented in Code Case N-318-1. Local stress levels in the pipe resulting from applied lug loads are obtained by multiplying the nominal st ress in the lug at the lug/pipe interface by the appropriate B or C index (as defined in Code Case N-318-1) for each individual lo ading condition. The local stresses are superimposed upon the general pipe stress as determined from program ADLPIPE to establish the total stress level in the pipe for that

loading condition. Loading conditions required to be considered for Plant Normal, Plant Upset, Plant Emergency, and Plan t Faulted Operating Condition are defined (per appropriate Updated FSAR section), and total stress in the pipe is obtained from summing the stresses for each individual loading condition that must be considered.

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Page 13 Local stress levels determined using B indices are added to the general stress levels from ADLPIPE and this sum is compared against allowable limits to demonstrate structural integrity. For the pipe wall, local stress levels determined using C indices are added to the general stress levels from ADLPIPE, and this sum is compared against the allowable range of

stress (S h + S a). The terminal ends of these portions of piping are considered to originate at a point adjacent to the restraints located inside and outside containment

which are: (a) Located reasonably close to the isolation valve (b) Capable of withstanding the loadings resulting from a postulated pipe rupture beyond this portion of the piping, such that neither valve operability nor the leaktight integrity of the containment is impaired. Details of typical containment pipi ng penetrations showing location of process pipe welds, anchorage and points of discontinuity are shown in

Figure 3.6(B)-3 and Figure 3.6(B)-4. Augmented in-service inspection of piping in containment penetration areas, along with inservice inspection of Code Class 2 components, is discussed in Section 6.6. 5. With the exception of piping penetrating Containment (Section 4. above) leakage cracks are postulated in ASME Section III Code Class 1 piping where the stress range by Equation 10 of Paragraph NB-3653 exceeds

1.2S m and in Class 2 and 3 or nonsafet y class piping where the stress by the sum of Equation (9) and (10) of Paragraph NC/ND-3652 exceeds

0.4(1.2 S h + S A). Nonsafety class piping which has not been evaluated to obtain similar stress information has cracks postulated at locations that result in the most severe environmental consequence.

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Page 14 b. Moderate Energy Piping Through-wall leakage cracks are postulated to occur in seismic Category I and nonnuclear fluid system piping located within or outside and adjacent to protective structures with the following exceptions: 1. Fluid system piping between isolation valves, provided they meet the requirements of ASME Section III, subarticle NE-1120, and are designed so that the maximum stress range does not exceed 0.4 (1.2 S h + S A) for ASME Class 2 piping. 2. Fluid system piping located in an area in which a break in a high energy system is postulated, provided a break in a moderate energy fluid system does not result in a more limiting condition than the break in the high energy system. 3. Seismic Category I fluid systems in which the maximum stress range in Class 2 or Class 3 or nonnuclear piping is less than 0.4 (1.2 S h + S A). Moderate and high energy piping cracks per a.5 above were postulated to occur in those locations that result in the maximum effects from spraying and/or flooding. Through-wall leakage cracks were postulated instead of breaks in the piping of those systems that qualify as high energy fluid systems for only short operational periods, but qualify as moderate energy fluid systems for the major operation period. These systems include containment spray, safety injection and residual heat removal. An operational period is considered short if the fraction of time that the system operates within the pressure-temperature limits specified for a high energy system is 2 percent or less of the time that it operates as a moderate energy fluid system.

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Page 15 c. Type of Break The following types of breaks and cracks were postulated to occur in high energy and moderate energy piping as described below:

1. High Energy Piping (a) Circumferential breaks were postulated to occur in high-energy piping larger than one inch nominal pipe size. Circumferential breaks are presumed to occur at right angles to the axis of the pipe, to completely sever the pipe within one millisecond and to separate the ends of the pipe to permit a fl ow area equal to the flow area of the pipe, except where pipe whip restraints function to limit pipe

separation. See Subsection 3.6(N).2.1 for exception for RCS piping. b) Longitudinal splits were postula ted to occur in high-energy piping four inches or larger nominal pipe size. The area of the longitudinal split was assumed to be equal to the flow area of the pipe, and the split was assumed to be parallel to the axis of the pipe. Jet impingement analysis was based on a rectangular break

2D i long by D i/8 wide where D i = pipe inside diameter. Breaks were oriented (but not concurrently) at two diametrically opposed points on the piping circumference such that the jet reactions

produce out-of-plane bending of the piping configuration. (c) Certain longitudinal break orient ations were excluded on the basis of the state of stress at the loca tion considered. Specifically, where the maximum stress range in the axia l direction is at least one and a half times that in the circumferential direction considering upset plant conditions, then only a circumferential break was postulated. (d) Longitudinal breaks were not pos tulated to occur in piping at terminal ends.

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Page 16 2. Moderate Energy Piping Through-wall leakage cracks were postulated to occur in moderate energy piping larger than one inch nominal pipe diameter, and to have openings of one-half pipe diameter by one-half the pipe wall thickness.

d. Jet Impingement Force Criteria The criteria used to evaluate jet impingement forces are described in Appendix 3C, Procedure for Evaluation of Jet Impingement Loads from High Energy Piping Failures. After jet forces imposed on structures or equipment have been determined, the capacity of the structures or equipment to support these loads without damage is investigated using conservative methods. Jet impingement loads are considered to be faulted condition loads and are so evaluated.

3.6(B).2.2 Analytical Methods to Define Forcing Functions and Response Models This section presents a description of the methods used to define forcing functions and response models for pipe whip analysis. For RC Loop piping, see Subsection 3.6(N).2.2.

a. Forcing Functions
1. Time Dependence The normal steady-state operating conditions of the plant were assumed prior to postulating a pipe rupture. When circumferential ruptures were postulated, the through-wall crack was assumed to develop across the circumference of the pipe instantaneously, and the ruptured pipe was assumed to separate to the full flow area (e.g., double ended rupture) in one millisecond. When longitudinal ruptures were postulated, the time for a longitudinal rupture to open to its maximum area was assumed to be one millisecond.
2. System Friction Loss Dependence In calculating forces acting on the piping system, full credit may be taken for any restrictions or line losses between the break and the pressure

reservoir(s).

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Page 17 3. Closed-Ended Lines For the closed end of a line (dead end or normally-closed valves) when it was obvious that the fluid dynamic for ces could not be sustained, pipe whip response was not considered.

4. Discharge Coefficient For flashing or nonflashing flow through circumferential and longitudinal breaks, a discharge coefficient, C d, of 1.0 was used to determine the flow rate through the break, Q = C d AV where: Q = flow rate through break A = break flow area V = velocity

C d = discharge coefficient

5. Options The jet thrust reaction, forcing function at the break locations may be generated from a dynamic fluid system model. However, a simplified approach was used, applying a maximum thrust value defined for discharge of nonflashing liquid or for discharge of saturated or superheated vapor as: T = k (P o - P a) A S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 18 where: T = jet thrust reaction force k = thrust factor (1.26 for steam and flashing fluids and 2.0 for sub-cooled, nonflashing fluids)

P o = fluid pressure in pipe (psig)

P a = ambient pressure outside system (psig) A = break flow area Limited pipe displacement at the break location, line restrictions, positive pump-controlled flow and the effects of pipe friction may be taken into account to reduce the jet force. For circumferential breaks, direction of thrust was assumed to be along the centerline of the pipe in a direction opposite the jet flow. For longitudinal breaks, thrust was assumed in a direction opposite jet flow. For all breaks, maximum thrust was assumed to occur within 1 millisecond and to be a steady-state condition thereafter.

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Page 19 3.6(B).2.3 Dynamic Analysis Methods to Verify Integrity and Operability

a. Dynamic Analysis Methods The analysis of a piping system and its restraints under pipe rupture conditions requires consideration of the interaction e ffects of both piping and restraints. The magnitude and distribution of loadings depend upon such parameters as the restraint load-deflection, gaps between piping and restraint, piping flexibility, break location, etc.
1. Energy-Balance Analysis In this method, kinetic energy gene rated during the first quarter cycle movement of the ruptured pipe is imparted to the piping/restraint system through impact and is converted into equivalent strain energy. Deformations of the pipe and the restraint are compatible with the level of absorbed energy. For applications where pipe rebound may occur upon impact of the restraint, an additional amplification factor 1.1 was used to establish the magnitude of the forcing function to determine the maximum reaction force of the restraint afte r the first quarter cycle response. Amplification factors other than 1.1 may be used if justified by more detailed dynamic analysis. Appendix 3D presents the procedure used for calculating piping/restraint system loads by the energy balance method.
2. Quasi-Static Analysis In order to satisfy the system capability requirements, a dynamic analysis is the preferred method. If dynamic analysis is not possible or feasible for a piping and restraint system, a quasi-static analysis may be possible if it is shown to give more conservative results. Two design considerations are required as in the dynamic analysis. The system must be capable of supporting both the dynamic and the

steady-state blowdown loads. If a constant, conservative blowdown force is assumed, the system is independent of the dynamic event occurrence time. Since the dynamic

inertia effects are therefore unknow n, the load-sharing relationship between the pipe and the restraints, etc., cannot be determined.

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Page 20 The jet force can be represented by a conservatively amplified static loading, and the ruptured system is analyzed statically. The amplification factor that is used to establish the magnitude of the forcing function is a conservative value obtained by comparison with the factors derived from detailed dynamic analysis performed on comparable systems. Appendix 3E presents the procedure used for calculating piping/restraint system loads by the equivalent static analysis method.

b. Design Considerations Pipe rupture locations and orientation were determined as stated in Subsections 3.6(B).2 and 3.6(N).2. Effects of each rupt ure were evaluated and, if necessary, whip restraints were located to protect the essential systems or components. For Code Class 1, 2 or 3 piping, the whip restraints were designed to prevent unrestrained whipping of the piping, but at the same time permit unrestrained thermal movement of the piping. In some cases, such as on the main steam and feedwater lines in the penetrations and piping tunnel areas, it was appropriate to use pipe whip restraint steel as an extension of the building steel for the attachment of seismic restraints. Wherever

this was done, the boundary between PWR steel and ASME Class 2 seismic restraints was defined by showing the PWR steel and the seismic restraints on

separate fabrication and installation dr awings. All Code Class supports are identified on the drawings as "N-Stamp Items." After the whip restraints were located, the following information was developed: Jet thrust force Pipe seismic displacement Pipe thermal displacement Maximum allowable pipe travel Insulation thickness S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Ru p ture of Pi p in g Revision 8 Section 3.6(B)

Page 21 Minimum gap between pipe and restraint is determined from consideration of 2, 3 and 5 above. Restraint stiffness is determined from 1 and 4. Where the whip restraint is also a seismic restraint, the following values for stiffness were used: For piping larger than 8" nominal diameter: 10 6 or 10 7 lb./inch For piping from 21/2" to 6" nominal diameter: 10 5 , 10 6 or 10 7 lb./inch For piping up to 2" nominal diameter: 10 4 , 10 5 or 10 6 lb./inch. Analyses of representative piping configur ations show that a change in stiffness of one order of magnitude in either direction will not change pipe stresses significantly, so that the designers generally used the lowest values for stiffness in the ranges given, unless pipe deflection is the critical parameter. In the design of the pipe whip restrain ts, the energy absorption capacity of the pipe was not considered. To determine the adequacy of a system , including pipes and restraints, following a postulated pipe rupture accident, two design considerations were evaluated:

1. Dynamic Response Upon the occurrence of the postulated pipe rupture, the system of pipe restraints, structure, etc., will respond dynamically to the suddenly applied blowdown thrust, F B(t). This thrust will move the pipe so that it impacts against the restraint with an impulse equal to the pipe mass times the impact velocity. The product of blowdown thrust, F B, and the time after this impact until motion ceases, t, will be an additional impulse on the system. 2. Static Equilibrium Following the occurrences of the dynamic event (when motion ceases), the system must be able to support the active applied forces (the blowdown thrust). Therefore, the system must satisfy the requirements of a static analysis.

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Page 22 For a conservative static analysis, each component (i.e., pipe, restraint) is capable of supporting the total load (or it is shown which component(s) support the load). When this is done and the components will have the load capacity to support the steady-state blowdown, the system design is considered to be conservative.

c. Piping Design Loading Combinations Pipes which have been identified in Subsection 3.6(B).1 as those which could cause adverse effects due to pipe movement were provided with means of controlling their motion, if barriers, separation, or some other acceptable method was not used for protection.
1. Adequacy Requirements To control the motion of pipes, it is necessary that the load on the pipe during the dynamic event be less than th e load capacity of the pipe. The dynamic load capacity of the pipe can be determined by test or by a suitable analytical model. Without testing, the load capacity for analysis is limited to the bending associated with a maximum fiber strain of 50 percent of the ultimate strain of the pipe material. Ultimate strain is defined as the value of strain which corresponds to the maximum stress on the engineering stress-strain diagram. For a given material where there is a range of values due to statistical variation, the guaranteed minimum value of ultimate strain is used. The second requirement, to insure that the motion of pipes is controlled, is for the moment-carrying capacity of the pipe to be greater than the applied moment after the occurrence of the dynamic event. An ultimate moment, M u, is defined as the maximum moment that the pipe cross section can support. If the applied moment, M a, defined as a force times lever arm is numerically smaller than M u, uncontrolled rotation of the pipe will not occur.

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Page 23 2. Material Properties Careful consideration was given to the piping material properties used.

The rapid loading conditions due to pipe rupture may require the consideration of high strain-rate effects on material behavior, in addition to strain-hardening considerations.Section III of the ASME Code provides tabulations of material properties which may be used for some evaluations. The values of yield strength at temperature, for example, are minimum values for static loadings. In calculating the allowable span distance between restraints, use of minimum values is conservative. In calculating the maximum moment which could be exerted on an anchor point, the use of minimum values woul d not be conservative. The applied moments, M a , during the steady-state blowdown will be no greater than 90 percent of the moment capacity of the pipe based on minimum pipe material properties determined from test, applicable specifications, or

codes. 3.6(B).2.4 Guard Pipe Assembly and Pipe Whip Restraint Design Criteria

a. Guard Pipe Assembly Guard pipes are used in the following locations: (a) on the main steam and feedwater lines to prevent pressurization of the containment enclosure in the event of a pipe rupture; (b) on the main steam lines just north of the main steam and feedwater pipe chase to protect the main steam isolation valves from damage due to jet impingement of the pipe chase north wall; and (c) on the main steam

line in the pipe bridge area to protect the Control Building wall from jet impingement. The guard pipes in the containment enclosure were designed as a

part of the flued head penetrations for the main steam and feedwater lines. A

discussion of the design criteria and analysis of the high energy containment penetrations is given in Reference 1. The purpose of the penetration assemblies is to permit penetration of the containmen t by process pipes without jeopardizing containment integrity. Where they are used as guard pipes, they also serve to prevent overpressurization of the containm ent enclosure and annulus. No other lines in this area require guard pipes.

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Page 24 In general, all process pipes penetrating containment are seamless. Penetration assemblies for large high temperature lines and steam generator blowdown lines are integrally forged flued head design. Penetration assemblies for cold lines or small lines (under 1" nominal diameter) are seamless pipe welded to flat plate heads which are in turn welded to sleeves anchored in the containment structure. All penetration sleeves are seal welded to the steel containment liner, and leak test channels are provided for periodic testing of containment leak-tightness. There are no process pipe welds located within the protective assemblies, with the exception of the 2" diameter steam blow down lines. The process pipe welds for these lines do not require in-service inspection (Reference IWC-1220d of ASME XI). Moment-limiting restraints have been pr ovided for all penetrations carrying high energy piping to maintain process pipe stress levels below the limits defined in Equation 8 of NC3652 for maximum stress range, considering all upset design transients in combination with OBE.

b. Pipe Whip Restraints For BOP piping, pipe whip restraints are provided where required to protect essential components, to maintain the motion of the ruptured pipe within controlled limits. The limit of pipe motion is the area within which no essential component would be affected by impact or jet impingement. The primary function of a pipe whip restraint is to control pipe motion upon the occurrence of a pipe rupture. As used in this context, a restraint is considered to be different from a pipe support. In certain instances, a restraint may also

function as a part of the building structure to support a pipe support, cable tray

support, conduit support, duct support or any other support or combination of these supports.

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Page 25 Typical whip restraints consist of h eavy structural extension of the building structure to the pipe, and a structural box or a series of U-bolts which surround the pipe to restrain lateral motion. Unle ss the restraint acts also as a support for a thermal or seismic restraint, contact betw een the pipe and the whip restraint is prevented by means of a suitable air gap. Where it is necessary to reduce pipe impact loads on critical structures, energy-absorbing devices are used between the pipe and the supporting structure. Pipe whip restraints are designed for one-time usage, and as such may be allowed to have greater distortion, plastic deformation, etc., than normally permitted for support design. In general, for pipe whip restraints, elastic design criteria were used. In cases where the pipe whip restraint was also used for attachment of supports or restraints, it was first designed and analyzed as a whip restraint, and then checked

to verify its ability to withstand the other support and restraint loads acting simultaneously. The basis for the designs and analyses of the pipe whip restraints is SRP 3.6.2 and 3.8.3, as referenced in SRP 6.2.1, paragr aph 6.2.1.2 addressing structures which behave linearly under the application of an applied load. If, as a result of pipe whip load reanalysis, the stress limits given above were exceeded, allowable stresses increased by faulted factors based on those allowed in ASME III, Division I, Appendix F, were applied. For some pipe whip restraints, certain members were designed to elasto-plastic design criteria. In those members designa ted to behave elasto-plastically, where the effects of strain-rate, strain-hardening, etc., are included, the permanent strain in these metallic ductile materials was limited to 50 percent of the uniform tensile strain. When a crushable energy-absorbing material was used, the deformation was limited to that corresponding to 75 percent of the crush pad core height. Where wire ropes were used as part of the pipe whip restraint design, Bethlehem Purple Plus IWRC wire rope material was selected, and design loads in the wire rope were limited to 55 percent of its material minimum breaking strength. Where energy absorption of the whipping pipe was dependent on elasto-plastic action of a pipe whip restraint member (U-bolts), the material selected exhibited high ultimate strain properties.

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Page 26 For the ASTM A193 GR-B8 material selected for the U-bolts, the ultimate tensile strain rate was limited to 40 percent. For the ASTM A276 Type XM-10 material selected for the U-bolts, the ultimate tensile strain rate was limited to 45 percent.

3.6.(B).2.5 Material to Be Submitted for the Operating License Review The summary of results of the analyses performed on high and moderate energy piping systems and their restraints to determine the effects of postulated pipe breaks and cracks, as well as the

procedures used, are presented in the following appendices:

Appendix 3A Pipe Break Analysis Summary Appendix 3C Procedure for Evaluating Jet Impingement Loads from High Energy Piping Failures Appendix 3D Procedure for Calculating Elasto-Plastically Designed Pipe Whip Restraint Loads by the Energy Balance Method Appendix 3E Procedure for Calculating Elastically-Designed Pipe Whip Restraint Loads by Equivalent Static Analysis Method.

3.6(B).3 References

1. "Stress Report for High Energy Piping Pe netrations for PSNH-Seabrook Station Units 1 and 2," Stress Report No. 9763-325-1, (Calculation 9763-C-01-ST-00-F) Rev. 1 dated September, 1976, United Engineers & Constructors Inc. 2. Moody, F.J., "Prediction of Blowdown Thrust and Jet Forces," Paper No.

69-HT-31, presented at the ASME-AIC HE Heat Transfer Conference, Minneapolis, Minnesota, August 3-6, 1969. 3. Final Rule Modifying 10 CFR 50 Appendix A, GDC-4 dated October 27, 1987

[52 FR 41288].

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Page 1 3.6(N) PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING This section describes the design bases and protective measures that are used to ensure the containment and its equipment are adequately protected against the dynamic effects caused by postulated rupture of the reactor coolant system piping. Elements of this system are discussed in Section 5.4. 3.6(N).1 Postulated Piping Failures in Fluid Systems outside of Containment Refer to Subsection 3.6(B).1. 3.6(N).2 Determination of Break Locations and Dynamic Effects Associated with Postulated Rupture of Piping The design bases for postulated breaks in the reactor coolant system piping are given below.

The design bases for the postulated pipe rupture in clude not only the break criteria, but also the criteria to protect other piping and vital systems from the effects of the postulated rupture.

3.6(N).2.1 Criteria Used to Determine Break and Crack Location and Configuration In any given piping system, there are a limited number of locations that are more susceptible to failure by virtue of stress or fatigue than the remainder of the system.

The discrete break locations and orientations in the reactor coolant loop were originally derived on the basis of stress and fatigue analysis. These postulated break locations and the methods that were used to determine them are described in Reference 1. An analysis of each individual reactor coolant loop confirmed the break locations defined in Reference 1. Actual seismic loads for the Seabrook site are included in the specific plant reactor coolant loop analysis. Eight of these break locations are eliminated from the plant design basis due to the consideration of the detailed fracture mechanics evaluation in Reference 5. This new design was approved by the NRC in Reference 4. For the remaining breaks, (a) pressurizer surge line nozzle, (b) residual heat removal line nozzle and (c) accumulator line nozzle, the breaks are postulated on terminal end criteria. At postulated circumferential break locations, the piping is assumed to separate and to allow double-ended flow. Longitudinal breaks are assumed to have an opening area equal to the flow

area of the pipe.

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Page 2 3.6(N).2.2 Analytical Methods to Define Forcing Functions and Response Models Following is a summary of the methods used to determine the dynamic response of the reactor coolant loop associated with postulated pipe breaks in the loop piping. a. Time Functions of Jet Thrust Force on Ruptured and Intact Loop Piping To determine the thrust and reactive fo rce loads to be applied to the reactor coolant loop during the postulated lo ss-of-coolant accident (LOCA), it is necessary to have a detailed description of the hydraulic transient. Hydraulic forcing functions are calculated for the intact reactor coolant loops as a result of a postulated LOCA. These forces result from the transient flow and pressure histories in the Reactor Coolant System. The calculation is performed in two steps. The first step is to calculate the transient pressure, mass flow rates, and thermodynamic properties as a function of tim

e. The second step uses the results obtained from the hydraulic analysis, al ong with input of areas and direction coordinates, and calculates the time history of forces at appropriate locations (e.g.

elbows) in the reactor coolant loops. The hydraulic model represents the behavior of the coolant fluid within the entire Reactor Coolant System. Key parameters calculated by the hydraulic model are pressure, mass flow rate, and density. Th ese are supplied to the thrust calculation, together with plant layout information, to determine the time-dependent loads exerted by the fluid on the loops. In ev aluating the hydraulic forcing functions during a postulated LOCA, the pressure and momentum flux terms are dominant. The inertia and gravitational terms are ta ken into account in evaluation of the local fluid conditions in the hydraulic model. The blowdown hydraulic analysis is required to provide the basic information concerning the dynamic behavior of the reactor core environment for the loop forces, reactor kinetics and core cooling analysis. This requires the ability to

predict the flow, quality, and pressure of the fluid throughout the reactor system.

The MULTIFLEX code (Reference 2) was developed with a capability to provide this information. The MULTIFLEX code performs a comprehensive space-time dependent analysis of a LOCA and is designed to treat all phases of the blowdown. The stages are: (1) a subcooled

stage where the rapidly changing pressure gradie nts in the subcooled fluid exert an influence upon the Reactor Coolant System and support struct ures, (2) a two phase depressurization stage, and (3) the saturated stage. See section 3.9(N).2.5 for details of the MULTIFLEX code.

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Page 3 The STHRUST computer program was developed to compute the transient (blowdown) hydraulic loads resulting from a LOCA. The blowdown hydraulic loads on primary loop components are computed from the equation 144 A gm14.7p144AF m 2 c 2 The symbols and units are: F = Force, lb.

A = Area, ft 2 P = System pressure, psia m = Mass flow rate, lb m/sec = Density, lb m/ft 3 g c = Gravitational constant = 32.174 ft-lb m/lb f - sec 2 A m = Mass flow area, ft 2 In the model to compute forcing functions, the Reactor Coolant Loop System is represented by a similar model as employe d in the blowdown analysis. The entire loop layout is represented in a global coordinate system. Each mode is fully described by: (1) blowdown hydraulic information and (2) the orientation of the streamlines of the force nodes in the system, which includes flow areas and projection coefficients along the three axes of the global coordinate system. Each node is modeled as a separate control volume, with one of two flow apertures associated with it. Two apertures are used to simulate a change in flow direction and area. Each force is divided into its x, y, and z components using the projection coefficients. The force components are then summed over the total number of apertures in any one node to give a total x force, total y force, and total z force. These thrust forces serve as input to the piping/restraint dynamic

analysis.

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Page 4 The STHRUST code is described in Reference 3. b. Dynamic Analysis of the Reactor Coolant Loop Equipment Supports The dynamic analysis of the reactor cool ant loop piping for the LOCA loadings is described in Section 3.9(N). 3.6(N).2.3 Dynamic Analysis Methods to Verify Integrity and Operability

a. Protective Measures The fluid discharged from the ruptur ed piping will produce thrust and reaction forces in the piping systems. The effects of these loadings are considered in assuring the continued integrity of the vital components and the Engineered Safety Features. To account for these effects in the design, a combination of component restraints, barriers, and layout is used to ensure that for a loss of coolant, steam or feedwater line break, propagation of damage from the original event is limited, and the components as needed, are protected and available. b. Criteria for Protection Against Postulated Pipe Breaks in Reactor Coolant System Piping A loss of reactor coolant accident is assumed to occur for a branch line break down to the restraint of the second normally open automatic isolation valve (Case II in Figure 3.6(N)-1) on outgoing lines and down to and including the second check valve (Case III in Figure 3.6(N)-1) on incoming lines normally with flow. (It is assumed that motion of the unsuppor ted line containing the isolation valves could cause failure of the operators of both valves to function.) A pipe break beyond the restraint or second check valv e will not result in an uncontrolled loss of reactor coolant if either of the two va lves in the line close. Accordingly, both of the automatic isolation valves are suitably protected and restrained as close to

the valves as possible so that a pipe break beyond the restraint will not jeopardize the integrity and operability of the valves. Further, periodic testing capability of the valves to perform their intended function is essential. This criterion takes

credit for only one of the two valves performing its intended function. For normally closed isolation or incoming check valves (Cases I and IV in Figure 3.6(N)-1), a loss of reactor coolant accident is assumed to occur for pipe breaks on the reactor side of the valve.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Revision 10 Section 3.6(N)

Page 5 Branch lines connected to the Reactor Coolant System are defined as "large" for the purpose of these criteria if they have an inside diameter greater than 4 inches up to the largest connecting line, generally the pressurizer surge line. Rupture of these lines results in a rapid blowdown from the Reactor Coolant System, and

protection is basically provided by the accumulators and the low head safety injection pumps (residual heat removal pumps). Branch lines connected to the Reactor Coolant System are defined as "small" if they have an inside diameter equal to or less than 4 inches. This size is such that emergency core cooling system analyses using realistic assumptions show that no clad damage is expected for a break area of up to 12.5 square inches corresponding to 4-inch inside diameter piping. Engineered Safety Features are provide d for core cooling and boration, pressure reduction and activity confinement in the event of a loss of reactor coolant or steam or feedwater line break accident to ensure that the public is protected in accordance with 10 CFR 100 guidelines. These safety systems have been designed to provide protection for a reactor coolant system pipe rupture of a size up to and including a double-ended severence of the reactor coolant system main loop. To ensure the continued integrity of the vital components and the Engineered Safety Systems, consideration is given to the consequential effects of the pipe break itself to the extent that: 1. The minimum performance capabilities of the Engineered Safety Systems are not reduced below that required to protect against the postulated break; 2. The containment (defined as the containment structure liner and penetrations, and the steam generator shell, the steam generator steam side instrumentation connections, the steam, feedwater, blowdown and steam generator drain pipes within the containment structure) leaktightness is not decreased below the design value, if the break leads to a loss of reactor coolant; and 3. Propagation of damage is limited in type and/or degree to the extent that: - A pipe break which is not a loss of reactor coolant will not cause a loss of reactor coolant or steam or feedwater line break.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Revision 10 Section 3.6(N)

Page 6 - A reactor coolant system pipe break will not cause a steam or feedwater system pipe break and vice versa. (a) Large Branch Lines Large branch line piping, as defined in Subsection 3.6(N).2.3b, is restrained to meet the following criteria in addition to Items 1 through 3 of Subsection 3.6(N).2.3b fo r a pipe break resulting in a loss of reactor coolant: (1) Propagation of the break to the unaffected loops is prevented to assure the delivery capacity of the accumulators and low head pumps. (2) Propagation of the break in the affected loop is permitted to occur but does not exceed 20 per cent of the flow area of the line which initially ruptured. This criterion has been

voluntarily applied so as not to substantially increase the

severity of the loss of coolant. (b) Small Branch Lines In the unlikely event that one of the small pressurized lines, as defined in Subsection 3.6(N).2.3b, should fail and result in a loss-of-coolant accident, the piping is restrained or arranged to meet the following criteria in addition to Items 1 through 3 of

Subsection 3.6(N).2.3b. (1) Break propagation is limited to the affected leg, i.e., propagation to the other leg of the affected loop and to the other loops is prevented. (2) Propagation of the break in the affected leg is permitted, but is limited to a total break area of 12.5 square inches, (4-inch inside diameter). The exception to this case is when the initiating small break is a cold leg high head safety injection line. Further propagation is not permitted

for this case.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Revision 10 Section 3.6(N)

Page 7 (3) Damage to the high head safety injection lines connected to the other leg of the affected loop or to the other loop is prevented. (4) Propagation of the break to a high head safety injection line connected to the affected leg is prevented if the line break results in a loss of core cooling capability due to a spilling injection line. c. Protective Provisions for Vital Equipment In addition to pipe restraints, barriers and layout are used to provide protection from pipe whip and blowdown jet forces. Some barriers used for protection against pi pe whip are: the crane wall serves as a barrier between the reactor coolant loops and the containment liner; in addition, the refueling cavity walls, the operating floor, and the crane wall enclose each reactor coolant loop into a separate compartment, thereby preventing an accident, which may occur in any loop, from affecting another loop or the containment liner; the portion of the steam and feedwater lines within the containment has been routed behind barriers that separate these lines from all reactor coolant

piping. The barriers described above will withstand loadings caused by jet forces and pipe whip impact forces. Other than for the emergency core cooling system lines that must circulate water to the vessel, the Engineered Safety Featur es are located outside of the crane wall. The emergency core cooling system lin es which penetrate the crane wall are routed around and outside the crane wall to penetrate the crane wall in the vicinity of the loop to which they are attached. It has been demonstrated by Westinghouse Nuclear Energy System tests, that lines hitting equal or larger size lines of the same schedule will not cause failure of the line being hit; e.g., a 1-inch line, should it fail, will not cause subsequent failure of a 1-inch or larger size line. The reverse, however, is assumed to be probable; i.e., a 4-inch line, should it fail and whip as a result of the fluid discharged through the line, could break smaller size lines such as neighboring 3-inch or 2-inch lines. In this cas e, the total break area is less than 12.5 square inches.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Revision 10 Section 3.6(N)

Page 8 Where restraints on the lines are necessary to prevent impact on and subsequent damage to the neighboring equipment or piping, restraint type and spacing are chosen so that a plastic hinge of the pipe at the two support points closest to the break is not formed. Alternatively, if the layout is planned so that whipping of

the two free sections cannot reach equipm ent or other pipes for which protection is required, then plastic hinge formation is allowed. As another alternative, barriers are erected to prevent the whipping pipe from impacting on equipment or piping requiri ng protection. Finally, tests and/or analyses are performed to demonstrate that the whipping pipe will not cause damage in excess of acceptable limits. Whipping in bending of a broken stainless steel pipe section does not cause this section to become a missile. This design basis has been demonstrated by Westinghouse Nuclear Energy Systems bending tests on large and small diameter, heavy and thin-walled stainless steel pipes. d. Design Loading Combinations As described in Section 3.9(N), the for ces associated with rupture of piping systems are considered in combination with normal operating loads and

earthquake loads for the design of supports and restraints to assure continued integrity of vital components and Engineered Safety Features. The stress limits for reactor coolant piping and supports are discussed in Section 3.9(N). 3.6(N).2.4 Guard Pipe Assembly Design Criteria Refer to Subsection 3.6(B).2.4. 3.6(N).2.5 Material to be Submitted at the Operating License Review

a. Table 3.6(N)-1 and Figure 3.6(N)-2 identif y the design basis pipe break locations for the main reactor coolant loop. The primary-plus-secondary stress intensity ranges and the fatigue cumulative usage factors at the design break locations specified in Subsection 3.6(B).2.1 are not tabulated since selection of these terminal end locations is independent of detailed stress and fatigue analysis.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Revision 10 Section 3.6(N)

Page 9 b. Design loading combinations and applicable criteria for ASME Class 1 components and supports are provided in Subsection 3.6(N).2.3d. Pipe rupture loads include not only the jet thrust fo rces acting on the piping but also jet impingement loads on the primary equipment and supports.

3.6(N).3 References

1. "Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop," WCAP-8082-P-A, January 1975, (Proprietary) and WCAP-8172-A (Nonproprietary), January 1975. 2. K. Takeuchi, et al., "MULTIFLEX A FORTRAN-IV Computer Program for Analyzing Thermal Hydraulic-Structure System Dynamics," WCAP-8708-P-A, Westinghouse Proprietary Class 2/ WCAP-8709-A, NES Class 3 (Non-Proprietary, September 1977. 3. "Documentation of Selected Westinghouse Structural Analysis Computer Codes," WCAP-8252, April 1974. 4. Final Rule Modifying 10 CFR 50 A ppendix A, GDC-4 dated October 27, 1987

[52 FR 41288]. 5. "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Seabrook Units 1 and 2," WCAP-10567, June 1984 (Proprietary) and WCAP-10566 (Nonproprietary), June 1984.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 1 3.7(B) SEISMIC DESIGN 3.7(B).1 Seismic Input This subsection contains a discussion of the input criteria to be used for seismic design of the plant. Items included in the discussion are design response spectra and the basis for their selection, earthquake time-motion records and the basis for their selection, response spectra obtained from time-motion records, and recommended percentages of critical damping to be used for seismic analysis. 3.7(B).1.1 Design Response Spectra The design response spectra for horizontal and vertical motion, corr esponding to the SSE applicable to the site, are presented in Subsection 2.5.2. The spectra amplification ratios for various levels of damping used to establish the design response spectra are based upon values presented in Regulatory Guide 1.60. The OBE response spectra are obtained by multiplying the SSE spectra ordinates by one-half. The duration of the earthquake is estimated at 10 to 15 seconds. There are no existing earthquake records pertinent to the Seabrook Station site. 3.7(B).1.2 Design Time History The three components of artificial time-history motion corresponding to the SSE are shown on Figure 3.7(B)-1, Figure 3.7(B)-2, and Figure 3.7(B)-3. These figures show the two horizontal motions and the one vertical motion, respectively. The components of the artificial time-history corresponding to the OBE are obtained by multiplying the ordinates of Figure 3.7(B)-1, Figure 3.7(B)-2, and Figure 3.7(B)-3 by one-half (1

/2). The earthquake motion was generated by super-imposing sinusoidal waves of many frequencies. Phase angles of the sinusoidal waves are randomly chosen. The resulting wave form is then multiplied by a trapezoidal intensity function to cause the time variation of intensity. All artificial time histories used in the analyses are base-line corrected.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 2 Response spectra obtained from earthquake time history are shown on Figure 3.7(B)-4, Figure 3.7(B)-5, Figure 3.7(B)-6, Figure 3.7(B)-7, Figure 3.7(B)-8, Figure 3.7(B)-9, Figure 3.7(B)-10, Figure 3.7(B)-11, Figure 3.7(B)-12, Figure 3.7(B)-13, Figure 3.7(B)-14, Figure 3.7(B)-15, Figure 3.7(B)-16, Figure 3.7(B)-17, Figure 3.7(B)-18, Figure 3.7(B)-19 and Figure 3.7(B)-20. Included on each figure is the corresponding design response spectra discussed in Subsection 3.7(B).1.1. The response spectra obtained from the time-motion record envelope the design response spectra for the period range of 0.03 seconds to 4.0 seconds. The response spectra have been computed using a method based on the exact solution of the governing differential equation for a single degree of fr eedom oscillator with viscous da mping. To ensure that the response spectra are sufficiently accurate, they were calculated at a set of discrete values for period, T, forming a geometric progression, i.e., To, T o r, T o r 2 ,. . . T o rn-1 r = 1.02 T o = 0.03 seconds This ratio corresponds to a period interval varying from 0.0006 seconds at a period of 0.03 seconds to a period interval of 0.01 seconds at a period of 0.50 seconds. 3.7(B).1.3 Critical Damping Values The percentages of critical viscous damping used for the seismic analysis of Category I structures, systems, and components are based on recommendations presented in Regulatory Guide 1.61. These percentages, which account for stress level as well as type of construction or fabrication, are summarized in Table 3.7(B)-1. For seismic piping analysis, an alternative to Regulatory Guide 1.61 may be used. These values are shown graphically in Figure 3.7(B)-38.

For the Cable Raceway System, an alternative to Regulatory Guide 1.61 may be us ed. Critical damping levels may be a maximum of 20 percent for input acceleration levels of 0.35g and greater for OBE and SSE conditions. In cases where input accelerations are between 0.1g and 0.35g, the critical damping values may be interpolated between 7 percent and 20 percent, respectively.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 3 3.7(B).1.4 Supporting Media for Category I Structures All seismic Category I structures are founded on sound bedrock or on engineered backfill extending to sound bedrock. Engineered backfill was also placed around all seismic Category I structures. The bedrock at the site is uniform, competent, and nonfragmented. Engineering properties of the bedrock measured in both the field and the laboratory are presented in Subsection 2.5.4.2a. The engineered backfill consists of either fill concrete, backfill concrete, offsite borrow, tunnel cuttings, or sand-cement. Properties of the engineered backfill materials are described in Subsection 2.5.4.5. The type of engineered backfill used beneath all seismic Category I structures was fill concrete, except for safety-related electrical duct banks, five electrical manholes, and the service water pipes, which were founded on offsite borrow or tunnel cuttings, as shown in Table 2.5-20. Identification of the safety-related electrical manholes founded on offsite bo rrow or tunnel cuttings, the depths of offsite borrow or tunnel cuttings over the bedrock under these particular manholes, the widths of their structural foundations and the total structural height are summarized below:

Manhole Numbers Depths of Soil over Bedrock (ft)

Widths of Structural Foundations (ft)

Total Structural Height (ft)

Supporting Material W13/W14 6-12 18 x 181/2 91/2 Offsite Borrow W15/W16 6-12 18 x 181/2 91/2 Offsite Borrow W19/20 15 231/2 x 231/2 12 Tunnel Cuttings W29/W30 14 19 x 221/2 15 Offsite Borrow W33/W34 18 18 x 181/2 12 Offsite Borrow All manholes are fully embedded.

The values of shear modulus, G, and shear wave velocity, v s , for both the offsite borrow and tunnel cuttings used for the analyses of the manholes ar e discussed in Subsection 2.5.4.7.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 4 3.7(B).2 Seismic System Analysis This subsection contains a discussion of the seismic analyses performed for seismic Category I structures and systems. Included in the discussion are the methods of seismic analysis used, the criteria used for mathematically modelling the structures and systems, the assumptions made in the analyses, and the effects considered. 3.7(B).2.1 Seismic Analysis Methods The seismic response of Category I structures, systems and components has been determined from suitable elastic dynamic analyses. The results of these analyses are used for the design of seismic Category I structures, systems and components, and are input for subsequent dynamic analyses. Two methods of seismic system analysis were used for seismic Category I structures: (1) the modal analysis response-spectrum method and (2) the mode-superposition time-history method. The time-history method was used to determine the dynamic response necessary to obtain amplified response spectra for component design.

The input forcing functions (the time history of ground motion) are shown graphically in Figure 3.7(B)-1, Figure 3.7(B)-2 and Figure 3.7(B)-3. The time history shown on Figure 3.7(B)-1 is used in both horizontal directions. The peak acceleration is 0.25g for the SSE and 0.125g for the OBE. Design response spectra for the response-spectrum method are shown in Section 2.5. The mathematical models used for the seismic Category I structures are typically lumped masses connected by linear elastic springs. Each struct ure, then, is described by a finite number of degrees-of-freedom chosen to represent the prin cipal overall behavior of the system. The modelling is described in Subsection 3.7(B).2.3 in more detail. The number of masses or degrees-of-freedom included in the analysis is determined by requiring the total degrees-of-freedom to exceed twice the number of significant modes with frequencies less than 33 Hz. Up to four degrees-of-freedom were considered for each mass point, three translation and one torsion. The three orthogonal di rections were run separately, and results were combined by the grouping method in accordance with Regulatory Guide 1.92. All significant modes with frequencies up to 50 Hz were used in analyses for both local and overall effects.

The effects due to inertial characteristics of flui d contained within a structural component were considered in the analysis by techniques described in Reference 1. No soil-structure interaction effects were involved becau se of the rock siting.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 5 The lumped mass mathematical models for representative seismic Category I structures, including the containment, are shown in Figure 3.7(B)-21, Figure 3.7(B)-22, Figure 3.7(B)-23, Figure 3.7(B)-24 and Figure 3.7(B)-25. Subsection 3.7(B)2.3 describes these models in more detail. Table 3.7(B)-2, Table 3.7(B)-3, Table 3.7(B)-4, Table 3.7(B)-5, Table 3.7(B)-6, Table 3.7(B)-7 and Table 3.7(B)-8 are the pr operty tables for the indicated models. Relative displacements between supports of structures are zero due to the base fixity. Maximum displacements throughout the structur e (relative to the fixed base of the model) were computed for use in component analysis, as described in Section 3.7(B).3. The description of the analysis methods used for seismic Category I systems and components is provided in other sections. 3.7(B).2.2 Natural Frequencies and Response Loads The seismic analyses of Category I struct ures are based upon the time history modal superposition and response spectra normal mode methods using idealized lumped mass models of the individual structures. These methods of an alyses use the natural fr equencies, mode shapes and appropriate damping coefficients of the system. The system frequencies, modes shapes and structural response are obtained using the STARDYNE computer program and the models described in Subsections 3.7(B).2.1 and 3.7(B).2.3.

Numerical results are tabulated in this subsection for the following representative seismic Category I structures: Containment Building, Primary Auxiliary Building, Control and Diesel Generator Building, and Fuel Storage Building. Table 3.7(B)-5, Table 3.7(B)-6, Table 3.7(B)-9 Table 3.7(B)-10, Table 3.7(B)-11, Table 3.7(B)-12, Table 3.7(B)-13, Table 3.7(B)-14, Table 3.7(B)-15, Table 3.7(B)-16, Table 3.7(B)-17, Table 3.7(B)-18, Table 3.7(B)-19, and Figure 3.7(B)-

20 list natural frequencies, periods, mode classification, nodal accelerations, nodal displacements and peak forces for these buildings.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 6 3.7(B).2.3 Procedures Used for Analytical Modeling The seismic analyses of Category I structures are based upon dynamic analyses using idealized three-dimensional lumped mass models of the physical structures. The inertial properties of the models are characterized by the lumped mass, eccentricity of the mass, and the torsional mass moment of inertia at each mass point of the model. The locations of lumped masses or nodes are selected at floor slabs, at cha nges of cross-sectional area and at intermediate points, such as locations of equipment, etc. The number of dynamic degrees-of-freedom are chosen to exceed twice the number of significant modes with frequencies less than 33Hz as a minimum. The concentrated masses are connected by weightless elastic beams representing the resisting structural members between mass points. Torsion is accounted for by rigidly offsetting the centers of mass from the centers of rigidity. Floor slabs are assumed to be rigid in their own plane, and the stiffness of shear walls is approp riately considered taking openings into account. The stiffness properties of the models are characterized by the cross-sectional area, moment of inertia, shear shape factor, torsional constant, and Young's Modulus of Elasticity and Poisson's Ratio. The torsional stiffness of the structure is evaluated using the lateral stiffness of the resisting elements between mass points and the square of their distance from the center of resistance for a particular story.

The stiffnesses of a particular story in the model are combined into a pseudo-elastic beam located at the center of resistance of the resisting elements. All seismic Category I structures, with the exception of some electrical manholes and ductbanks (see Subsection 3.7(B).2.4), are supported on competent bedrock or concrete fill over bedrock. All mathematical models are, therefore, fixed agai nst translation and rotati on at their bases. The elevation of the point-of-fixity of the mathematical model is determined as follows: a. Lowest elevation of upper surface of concrete backfill which bears directly against the structure (i.e., no seismic isolation material), or b. Elevation of a continuous floor slab which does not amplify the ground response, whichever is higher. The actual stiffness of resisting elements that extend below the assumed point-of-fixity is appropriately considered and reflected in the calculation of pseudo beam properties. Equipment having relatively small mass/or high frequency is decoupled from the supporting structure, but its mass is included with the mass of the supporting system. The NSSS, a major equipment system whose dynamic behavior can have important interaction with the supporting containment concrete internals structure, was mo deled as a coupled syst em, NSSS and structure, by Westinghouse in their dynamic analyses. The struct ural portion of the m ode is represented by Figure 3.7(B)-22.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 7 The hydrodynamic mass effects of the fluid inside the structure are consid ered in modeling the inertia properties. Equivalent we ights of fluid that are effective in producing impulsive forces and the constrained weight of fluid are determined and included in the lumped mass model at the appropriate nodes. These weights were calculated using the method described in Reference 1. The combined effects of the two horizontal and one vertical motion are taken into account to obtain the design parameters. The three seismic re sponses or effects at a particular point caused by each of the three orthogonal components of seismic motion are combined by taking the square-root-of-the-sum-of-the-squares of the particular effect or response at that point, in accordance with Regulatory Guide 1.92. For symmetric structures, such as the containment, coupling between translational and torsional modes of vibration is negligible, and the dynamic degrees of freedom are therefore uncoupled for the two horizontal directions. The seismic analysis of structures in the vertical direction is performed separately considering different mathematical models of the structures. For this analysis, only the vertical dynamic degrees-of-freedom associated with each mass are retained.

Ground response spectra for the Seabrook site (Subsection 3.7(B)

.1) are in accordance with Regulatory Guide 1.60, factored by 0.25 to match the site maximum acceleration. These spectra are valid for structure and component design provided the system in question is supported on bedrock or concrete fill over bedrock, or is a por tion of a structure which is subjected to ground response. The ground response spectra or artificial time-histories are a pplied as input at the structure's point-of-fixity. The lumped-mass models of Figure 3.7(B)-21, Figure 3.7(B)-22, Figure 3.7(B)-23, Figure 3.7(B)-24 and Figure 3.7(B)-25 have the properties listed in Table 3.7(B)-2, Table 3.7(B)-3, Table 3.7(B)-4, Table 3.7(B)-5, Table 3.7(B)-6, Table 3.7(B)-7 and Figure 3.7(B)-8, as noted pr eviously in Subsection 3.7(B)2.1. 3.7(B).2.4 Soil/Structure Interaction All major seismic Category I structures are founde d on rock or concrete extending to rock, thus permitting a fixed base approach to be used. Static and dynamic earth pressures for Category I structures surrounded by offsite borrow are discussed in Subsection 2.5.4.10c, and the pressure diagrams are shown on Figure 2.5-66 and Figure 2.5-67. Analysis methods for computing static and dynamic earth pressures on seismic Category I structures are provided in Subsection 2.5.4.10c. There are no hydrodynamic effects due to groundwat er because there are no structures exposed to free groundwater. The only dynamic effect produced by the pr esence of ground water is an increase in dynamic lateral soil pressure resulting from an increase in density of backfill material from that of the moist condition to that of the saturated condition.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 8 Certain seismic Category I electrical manholes are founded on a thin stratum of soil (maximum depth of soil between the foundation and the bedrock is 18 feet), as described in Subsection 3.7(B).1.4. These manholes are analyzed using the multiple lumped mass-spring approach described by Whitman in Reference 2. All seismic Category I manholes are completely embedded. The effect of embedment is to increase the soil spring stiffness thus increasing the natural frequency of the system resulting in reduced seismic design value. Hence, it was conservatively assumed to neglect embedment effect. The input design gr ound response spectra, discussed in Subsection 3.7(B).1.1, which were used for the analysis indicated that the in crease in frequency would result in decrease in response. To reduce amplification properties of the soil between the ground surface and the rock, the backfill material is controlled by controlling the placement requirement and material characteristics. The soil property given in Updated FSAR Subsection 3.7(B).1.4 is used for determining soil stiffness. The variation in soil parameters was not considered for the analysis. The lowest shear wave velocity was used to obtain lowest structural freq uency which would give higher structural respons es as stated above. The above conservative assumption combined with lower system damping values of 7 percent for SSE and 4 percent for OBE causes seismic load in excess of the actual value.

The design of manholes is governed by dynamic so il pressure; the inertia loads had a minimum effect on the design.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 9 3.7(B).2.5 Development of Floor Response Spectra The analysis for overall dynamic response of the structural system is described in Subsections 3.7(B).2.1 through 3.7(B)2.3. Local amplification of overall response (generally of floor slabs in the vertical direction) is represented by amplified response spectra (ARS). A time-history seismic analysis is used to generate these spectra using either of two methods. In the first method, the local amplification of slabs, beam and columns is evaluated and an appropriate range of frequencies selected for all local frequencies below 33 Hz. Single-degree-of-freedom (SDOF)' systems representing the computed range of local frequenc ies are connected to the stick model of the overall structure such that the total weight at the elevation in question (i.e., summation of the weight of the SDOF nodes plus the stick model node) equals the total weight, as if the slabs, beams, columns and walls were all rigid. The total stick model is then analyzed using the ground motion artificial time histories described in Subsection 3.7(B).1 as the input forcing function. The second method of evaluation of local amplification is to model the sub-structure with finite elements in sufficient detail to predict local modes of vibration. The response time-history from the overall stick model, at the elevation of the sub-structure, is then used as the input forcing function. For symmetric structures, separate analyses are performed for two horizont al directions, and the individual time-history motions and amplified response spectra ar e obtained for the nodal point locations of the lumped-mass model. For unsy mmetric structures, the floor response spectra values in both horizontal direct ions are initially obtained due to each horizontal component of motion. The resultant response spectra for each horizontal direction are then obtained by combining co-directional responses according to the square-root-of-the-sum-of-the-squares method. The floor response spectra in the vertic al direction are obtained separately from the vertical response time-history motions. No coupling of vertical and horizontal motion exists, since rock foundation precludes rocking motion. Structural damping used in the seismic analysis of all Seabrook structures is in conformance with Regulatory Guide 1.61. Amplified response spectra are generated at 1, 2 and 4 percent equipment damping associated with the OBE, and at 1, 2, 3, 4 and 7 percent equipment damping for the SSE.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 10 To obtain complete and accurate response spectra with respect to peak values, the spectrum ordinates are calculated at natural frequencies of supporting structures where peaks are normally expected, and at other frequencies at sufficiently small frequency intervals. Frequencies listed in Table 3.7(B)-21 are used to compute floor response spectra. This table is based on Table N-1226-1 of the ASME Boiler and Pressure Vessel Code,Section III, Division I, Nuclear Power Plant Components, 1980 Edition, Appendix N, 'Dynamic Analysis Methods.' The floor response spectra generated based on Table 3.7(B)-21 including the natu ral frequencies of supporting structures will meet the intent of SRP Subsection 3.7.1 (and also R.G. 1.122). The response pectra developed at the subsystem supports as described above, and widened and enveloped as described in Subsection 3.7(B).2.9, are used direct ly, for the analysis of the subsystem. See Figure 3.7(B)-26, Figure 3.7(B)-27, Figure 3.7(B)-28 and Figure 3.7(B)-29 for typical amplified response spectra for representative seismic Category I structures. Figure 3.7(B)-35 and Figure 3.7(B)-36 show typical comparisons of enveloped floor response spectra for 1 and 4 percent damping gene rated using frequency intervals based on Table 3.7(B)-21 and R.G. 1.122. A single absolute maximum spectrum is generated from the appropriate response points if more than one set of spectra are generated for a particular elevation or portion of a floor area. 3.7(B).2.6 Three Components of Earthquake Motion The seismic analyses of the Category I structures are performed using three-dimensional lumped mass models of the structures. The maximum response in principal directi ons is calculated using each of the three components of earthquake motion, and then the responses in each direction are obtained by taking the square-root-of-the-sum-of-the-squares of the maximum co-directional responses at a particular point in the structure.

This procedure is in accordance with Standard Review Plan 3.7.2,Section II.6a and b(1). 3.7(B).2.7 Combination of Modal Responses Modal responses are combined as per Regulatory Guide 1.92, paragraph 1.1 and 1.2.1. When the response spectrum method of analysis is used to determine seismic response of seismic Category I structures, the most pr obable response is obtained as the square-root-of-the-sum-of-the-squares of the responses from the individual modes. For the closely spaced modes having frequencies within 10 percent of each other, the structural response is obtained by first obtaining the absolute sum of the re sponses of the closely spaced modes and then combining this sum with other remaining modal responses using the square-root-of-the-sum-of-the-squares method.

See Subsection 9.1.4.3.a.5(d)(1) for the methodology used in support of upgrading the Cask Handling Crane to meet the single failure proof requirements of NUREG-0554.

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Page 11 3.7(B).2.8 Interaction of Non-Category I St ructures with Category I Structures All nonseismic Category I structures which, due to their proximity to seismic Category I structures, could possibly compromi se the safety function of the seismic Category I structures by their collapse, are either designed for SSE loading or are designed to collapse away from the adjacent seismic Category I structures (see Table 3.7(B)-22 for further discussion). The methods described in Subsection 3.7(B).2.1 are used for the analyses of these structures. 3.7(B).2.9 Effects of Parameter Variation on Floor Response Spectra The location of the peak responses on amplified response spectra curves vary as a result of the variation in material properties. This impacts subsystem design as discussed in Subsection 3.7(B).3. The variability in structural materials' properties and modeling assumptions is accounted for by peak spreading when generating envelopes of the response spectra. All Category I structures for which in-structure response spectra are generated are supported on rock

and, hence, variability of soil properties is not th e consideration in broadening the peaks of floor response spectra. A value of at least +/- 10 percen t is used to broaden the peaks of the floor response spectra associated with the structural frequencies. Regulatory Guide 1.122 is complied with except as noted in Section 1.8. 3.7(B).2.10 Use of Constant Vertical Load Factors Detailed vertical seismic system analyses are performed for seismic Category I structures; therefore, constant vertical static factors are not used. 3.7(B).2.11 Method Used to Account for Torsional Effects Torsional degrees-of-freedom are included in the dynamic analysis to obtain the natural frequencies and mode shapes of seismic Category I structures. For structures in which the center of mass and center of rigi dity is not coincident, torsional effects are automatically considered when the response to horizontal motion is obta ined. The design story shears are computed by considering both the direct shear from the orthogonal earthquake components and the shear due to torque. An accidental torsion, based on 5 percent eccentricity (SRP 3.7.2, Rev. 1, July 1981) was shown to have negligible effect on the design.

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Page 12 3.7(B).2.12 Comparison of Responses The nodal displacement and acceleration responses obtained from response spectrum and time-history methods of analysis are compared for the Containment Building and Fuel Storage Building. Table 3.7(B)-10 and Table 3.7(B)-11 list the displacements and accelerations computed by both methods for horizontal and vertical directions and SSE and OBE conditions (as noted on the tables) for the Containment Build ing. Referenced elevations correspond to mass points as shown on Figure 3.7(B)-21. Table 3.7(B)-19 shows similar information for the Fuel Storage Building, whose model is shown on Figure 3.7(B)-25. Approximate equivalency can be seen in the results of the two methods. 3.7(B).2.13 Methods for Seismic Analysis of Category I Dams This section is not ap plicable to Seabrook. 3.7(B).2.14 Determination of Seismic Category I Structu re Overturning Moments All seismic Category I structures are designed to resist overturning due to the combined effects of the vertical and two horizontal components of seismic ground motion. A structure's ability to resist overturning is calculated by either of two conservative approaches: moment equilibrium or a work-kinetic energy method. a. Moment Equilibrium Method In the moment equilibrium method, the response of a structure due to three directions of earthquake is obtained from the dynamic analyses. The maximum overturning moment, about the toe of the mat, is computed as:

M O = M h1 or M h2 where: M O = overturning moment M h1 = maximum overturning moment in one horizontal direction due to effects of three directions of earthquake.

M h2 = maximum overturning moment in second horizontal direction due to eff ects of three direc tions of earthquake.

M h1 or M h2 include all dynamic effect s on the structure including the dynamic effect of soil caused by seismic motions.

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Page 13 The resisting moment is computed as follows:

M R = (W - V S) X 1 - (V h) X 2 + M b where: M R = resisting moment W = weight of the structure and any fill

V s = maximum vertical seismic fo rce of structure acting upward due to three directions of earthquake V h = maximum vertical hydrostatic force on the structure acting upward X 1 = horizontal distance of the centroid of the structure from the toe of the mat X 2 = horizontal distance to center of hydrostatic area from the toe of the mat M b = resisting moment due to key action of mat or passive resistance of the structure. The factor of safety against overturning is:

M M FS O R As long as the factor of safety is equal to or grea ter than 1.10 for SSE and 1.50 for OBE load conditions, the structure is c onsidered stable against overturning. b. Work-Kinetic Energy Method In the work-kinetic energy method, the ki netic energy in a st ructure due to an OBE or SSE is estimated by: 2 i V 2 i H iVV M 2 1 KE M i = Mass concentration at some point (i) in the structure (V H)i = Maximum total lateral velocity S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 14 (V v)i = Maximum total vertical velocity (V H)i and (V v)i are computed as follows: (V H)i 2 = (V x)i 2 + (V H)g 2 (V v)i 2 = (V z)i 2 + (V v)g 2 where: (V H)g = Peak horizontal ground velocity (V v)g = Peak vertical ground velocity (V x)i = Maximum relative lateral velocity of mass, M i, due to three directions of earthquake (V z)i = Maximum relative vertical velocity of mass, M i , due to three directions of earthquake. (V H)g and (V v)g are obtained from ground response spectra of the time history used in seismic analysis. (V x)i and (V z)i are obtained from seismic analysis of the structure. The work, W, required to overturn the structure is computed as: W = M t g h + W p - W b where:

M t = Total mass of the structure and foundation mat g = Gravitational acceleration h = The vertical distance for which the center of mass of a structure must be lifted to reach the overturning position W p = The additional work required to displace the soil on the toe side of an embedded structure W b = The work done by the buoyancy force on the submerged portion of a structure.

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Page 15 Each collinear effect (as described in both a. and

b. above) due to the eff ects of three directions of earthquake is obtained by combining the re spective individual components using the SRSS method. The structure is considered stable against overturning when the ratio W/KE exceeds the safety factors described in a. above. 3.7(B).2.15 Analysis Procedure for Damping When the components of a seismic Category I structure or system are c onstructed of different materials, and these components cannot be uncoupled due to dynamic interaction effects, an average modal damping value is used for the dynamic analysis of the system. This average modal value is computed from the damping valu es of the various com ponents, as listed in Table 3.7(B)-1, each weighted by the energy stored in these components in the various modes of vibration. This value is computed as:

EED D Tnini n where: D n = Average damping value of the n th mode D i = Damping value of the i th component E in = Energy stored in the i th component in the n th mode E Tn = Total energy stored in the structure or system in the n th mode. The energy values are computed from the stiffness matrices of the various components and the mode shapes associated with undamped vibration. By this procedure, a diagonal damping matrix is computed allowing for the uncoupling of the equations of motion and the use of modal super-position. 3.7(B).3 Seismic Subsystem Analysis This section describes the seismic analysis performed on all subsystems, exclusive of those within the NSSS supplier's scope of responsibility. See Subsection 3.7(N).3 for the NSSS supplier's discussion.

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Page 16 3.7(B).3.1 Seismic Analysis Methods The seismic analyses of all Category I subsystems and components use either a dynamic analysis method or an equivalent static load method. Selection of the particular method of analysis depends upon the suitability of the mathematical model to adequately represent the behavior of the system or component. All points of the systems or components where the deflections, loads and stresses are expected to be significant are included in the mathematical model. The use of the equivalent static load method is limited to systems which can be represented by simple mathematical models. a. Dynamic Analysis Dynamic analyses were performed on subsystems and components such as electrical cable tray supports, electrical conduit supports, instrument racks, battery racks, etc., using the modal response spectrum analysis technique. The mathematical model for the electrical cable tray supports, as represented by Figure 3.7(B)-30, is typically represented by Figure 3.7(B)-31. Modeling considerations are discussed in Subsection 3.7(B).3.3b. The number of modes considered in determining the response of the equipment or components is such that inclusion of additional modes would not result in more than a 10 percent increase in the response. b. Equivalent Static Analysis The equivalent static analysis met hod consists of applying a load at the center-of-gravity of the component or equipment in the direction of seismic excitation, where the applied load is calculated by multiplying the total weight of the component or equipment with the applicable seismic acceleration level corresponding to the fundamental frequency of the component or equipment, using an appropriate static coefficient to account for the combined modal participation of the higher modes of vibration. When the equipment or component fundamental frequency is less than 33 Hertz, a stat ic coefficient of 1.5 will be used, except where a reduced static coefficient can be justified. When the equipment or component fundamental fr equency is equal to or greater than 33 Hertz, a static coefficient of 1.0 will be used when the equipment or component has a single predominant frequency. When there is more than one predominant frequency, a static coefficient of 1.3 will be used, except where a reduced static coefficient can be justified. The criteria for selecting the applicable seismic acceleration level are as follows:

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Page 17 1. When the equipment or component fundamental frequency is greater than 33 Hertz, the acceleration level corresponds to the

'g' value of the rigid range of the amplified floor response spectra. 2. When the equipment or component fundamental frequency is less than or equal to 33 Hertz, or is not evaluated, the acceleration level corresponds to the peak 'g' value of the amplified floor response spectra. However, if the equipment or component fundamental frequency is predominant over all

other frequencies, the actual 'g' value corresponding to the fundamental

frequency is used, except that such 'g' value must be equal to or greater than the maximum 'g' values corres ponding to all higher frequencies. 3.7(B).3.2 Determination of Number of Earthquake Cycles During the forty-year design life of the plant, the occurrence of one safe shutdown earthquake (SSE) and five operating basis earthquakes (OBE) are postulate

d. For fatigue analysis requirements of safety Class 1 components, a maximum of twenty stress cycles per earthquake is postulated, which results in a total of one hundred cycles for the five OBE events. For the SSE, no estimate of the number of stress cycles is made, since it is a faulted condition, and a fatigue analysis is not required. 3.7(B).3.3 Procedure Used for Modeling
a. Modeling of Piping Systems for Dynamic Analysis The piping systems were analyzed using a three-dimensional structural model composed of concentrated lumped masses connected by massless spring elements having the same strength and stiffness properties as the pipe. The model accounts for the interaction effect between piping, equipment a nd supports. Supports were modeled as flexible members with the appropriate spring rate to represent the support stiffness. The piping model was terminated at equipment nozzles that were modeled as rigid anchors with consideration given to the seismic amplification of equipment, as follows: 1. For rigid equipment in which the fundamental frequency was equal or higher than 33 Hz, the amplified response spectra of the structure was used. 2. For equipment in which the fundamental frequency was lower than 33 Hz, the amplified response spectra and the seismic anchor displacement of the equipment at the pipe/nozzl e interface point was used.

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Page 18 All in-line components were included in the model. The concentrated mass of in-line components such as valves, flanges, and strainers were represented as lumped masses. Valve operators were modeled as an offset lumped mass to account for the torsional and in-pla ne bending effects on the piping. The following criteria were used for the decoupling of piping subsystems: 1. Piping was decoupled from the equipment, and the nozzle modeled as a full, six-degree-of-freedom restraint. 2. Branch connections were decoupled from the main runs when the ratio of the branch to run section moduli was equal to or less than 0.05. 3. Piping subsystems which were decoupled into separate analytical models satisfied one of the following criteria: (a) The boundary of the decoupling point is a full anchor for the piping of both separate models. (b) The boundary of each decoupled model contains a region of common overlap to both models wh ich provides restraint(s) in each of the three orthogonal directions. b. Modeling of Equipment Seismic Category I equipment is modeled as lumped-mass spring systems which consist of a series of discrete mass points connected by massless elastic members. All significant concentrated weights are represented as lumped masses. Typical examples of concentrated weights are weights of motor rotor and pump impeller in the analysis of shafts. In general, masses are lumped at points where the maximum displacements are anticipated to occur. The number of lumped masses are such that the mathematical model represents the equipment response as closely as possible. 3.7(B).3.4 Basis for Selection of Frequencies All frequencies below 33 Hz are considered in computing the total response of the equipment and components. Whenever possible, equipmen t and components are de signed so that their fundamental frequencies are less th an half or more than twice the dominant frequencies of the support structure. Where this is found to be impractical or impossible to achieve, the equipment and components are adequately designed for the amplified loading.

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Page 19 3.7(B).3.5 Use of Equivalent Stat ic Load Method of Analysis The criteria and procedur e for the use of the equivalent static load method are described in Subsection 3.7(B).3.1b. 3.7(B).3.6 Three Components of Earthquake Motion Seismic Category I subsystems and components are analyzed by considering the combined effects of seismic loads occurring in three mutually perpendicular directions, two in the horizontal direction and one in the vertical direction. The total combined response (displacements, stresses, and forces) due to the three components of earthquake motion is obtained by using the square root of the sum of the squares method as follows:

2 j 31j c X X Where X c = Total combined response of the parameter x (displacement, stress or force etc.) X j = Value of the combined response of the parameter x in the j-direction of the earthquake. 3.7(B).3.7 Procedures for Combining Modal Responses The combined response of equipment and com ponents in a given earthquake direction is obtained by combining the individual modal responses using the square-root-of-the-sum-of-the-squares method. For equipment and components having modes with closely spaced frequencies, this method is m odified to include the po ssible effects of these modes. The groups of closely spaced modes are chosen such that the difference between the frequencies of the consecutive modes in a group doe s not exceed ten percent. No one frequency is in more than one group. For closely spaced modes, the modal response in a particular earthquake direction is obtained by using Equation (1) for piping and associated in-line components and by Equation (2) for equipment and other components. For closely spaced modes, X j is given by:

2 1 2 mq nq1m P1q 2 k N'1k j XXX

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Page 20 Where, nNN'q P1q P = Number of groups of closely spaced modes n q = Number of closely spaced modes in group number q N - Total number of modes X mq = Maximum value of the response of the element attributed to the mth mode of group number q

X K = Maximum value of the response of the element due to Kth mode An alternate expression for X j is given by:

XXX m R1m 2 1 2 k N'1k j Where the variables are as defi ned above except for the following: N' = N-R R = Modes with closely spaced frequencies.

X m = Maximum value of the response of the element attributed to the mth mode. The combined response of piping in a given ea rthquake direction is obt ained by combining the individual modal respons es using the square-root-of-the-sum-of-the-squares (SRSS) method. The closely spaced mode technique described in Subsection 3.7(B).3.7 may also be used.

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Page 21 3.7(B).3.8 Analytical Pr ocedures for Piping The procedure used for modeling of piping systems is described in Subsection 3.7(B).3.3a. The analytical procedures applicable to piping systems are as follows: a. Dynamic Analysis of Piping Systems Detailed seismic analysis of piping systems is performed by using finite element analysis programs to ensure that the stresses in the piping system meet the applicable ASME Section III Code requirements. Modal Response Spectrum Analysis technique is used for the dynamic analysis of piping systems. The seismic loading input consists of amplif ied floor response spectra obtained for discrete structural locations. The piping system is represented by a three-dimensional lumped-mass model, and is analyzed by determining the response of the system to the three components of the earthquake motion. The procedure for determining the total combined response is described in Subsection 3.7(B).3.6. b. Procedures to Account for Differential Piping Support Movement The piping system is analyzed for differential support movement at different support points located within a building and between buildings by using the static analysis method, the maximum relative support displacements are obtained from structural response calculations, and the worst differential movements between support points is used for the piping analysis. Pipe Stresses from support movement analysis are combined with the other stresses in appropriate Code equations. Loads on supports are added to other applicable loads. c. Equivalent Static Analysis of Piping Systems When seismic analysis of piping systems is performed using the static analysis approach, the total combined response (displacements, stresses, and forces) due to the three components of earthquake motion is obtained by using the square-root-of-the-sum-of-the-squares method. The response of the above parameters in each earthquake direction is obtained by applying a uniformly distributed 'g' load in that direction multiplied by a static coefficient of 1.5. The magnitude of the 'g' value corresponds to the fundamental frequency of the system, and is obtained from the applicable amplified response spectra curve.

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Page 22 3.7(B).3.9 Multiply-Supported Equipment Components with Distinct Inputs When the response spectrum method is used to analyze equipment and components supported at different elevations within a building or betw een buildings, typically, an upper bound envelope of all the individual response spectra for these lo cations is used to calcula te the inertial responses of the multiply-supported equipment and components. In addition, the relative displacements at the support points are determined according to the procedure outlined in Subsection 3.7(B).3.8b. Where the response of selected items in the analytical model is under the direct influence of a specific support point amplified re sponse spectra, then this analysis may be performed using this amplified response spectra. Additionally, where necessary, alternate methods are used which employ acceptable techniques which consider multiple amplified response spectra. 3.7(B).3.10 Use of Constant V ertical Static Factors Constant vertical static factors are not used for the design of seismic Category I subsystems. 3.7(B).3.11 Torsional Effect of Eccentric Masses The mathematical model generated to represent a piping system incl udes consideration of eccentric masses, such as valve operators. The eccentric mass is represented in the model as a lumped mass at the end of a weightless beam cantilevered from the valves. The length of the beam is equal to the distance between the center of gravity of the operator and the axis of the pipe. 3.7(B).3.12 Buried Seismic Category I Piping Systems and Tunnels Seismic Category I, ASME Section III, piping systems have been analyzed by utilizing the schemes and procedures developed by Newmark and Rosenblueth (Reference 3), Shah and Chu (Reference 4) and Iqbal a nd Goodling (Reference 5). The criteria for analysis of buried piping systems is to demonstrate their capability to withstand seismic soil strain and internal pressure loads without exceeding code allowable stress levels.

Analysis of these underground pipes subjected to seismic ground mo tion is based on the piping system configuration, boundary conditions, and the elastic properties of the soil and piping. Seismic Category I, ASME Section III, buried piping which penetrates structures where seismic soil-structure differential movements are expected is protected by providing adequate clearance between pipe and sleeve at the penetrations or by additional bend for flexibility.

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Page 23 There are no seismic Category tunnels which are analyzed as subsystems. Electrical, piping and passage tunnels are analyzed as seismic Category I structures, a nd are discussed in Subsection 3.7.2. 3.7(B).3.13 Interaction of Other Piping with Seismic Category I Piping The loading effects of nonseismic Category piping portions on seismic Category I piping is kept to a minimum by providing a restraint, or a group of restraints, at or beyond the defined boundaries, and extending the seismic analysis performed for the seismic Category I portion of the piping system to the boundary re straints. These restraints are designed to withstand the most severe loading combinations that could result from the nonseismic Category I piping. More specifically, to assure that an earthquake of SSE intensity will not result in failure of seismic Category I piping, the mathematical models constructed for Category I piping include the nonseismic Category I piping up to a boundary restraint (defined as a set of one or more restraints or anchor). Typical functions an d locations of these re straints are shown in Figure 3.7(B)-37. In Case 1, moments and forces introduced by the nonseismic piping are reacted through the six degree-of-freedom restraint as shown. In Case 2, moments are reacted by couples between pairs of restraints which limit motion in the indicated directions. These boundary restraints are designed to withstand the plastic capabili ty of the contiguous piping.

The structural integrity of the nonseismic Category I piping is assured by comp liance with the stress limits shown on Figure 3.7(B)-37. Class 2 or 3 piping must satisfy the limits of ASME III, NC-3600 or ND-3600. Pipes and supports in th e shaded regions are included in the mathematical model of the seismic Category I piping and are subjected to SSE loads. Nonseismic piping beyond the boundary restraints satisfies the limits of B31.1. Loading conditions and stress limits for all pipe classi fications are summarized in Subsection 3.9.3. 3.7(B).3.14 Seismic Analysis of Reactor Internals See Subsection 3.7(N).3.14. 3.7(B).3.15 Analysis Procedure for Damping Where the equipment or component consists of subcomponents with the same damping characteristics, the same critic al damping value is used for the entire equipment or component. The corresponding critical damping value is chosen from Table 3.7(B)-23. For seismic Category I equipment or component consisting of subcomponents with different damping characteristics, the lowest critical damping value associated with the subcomponents in the equipment or component is used in the analysis for all modes.

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Page 24 Alternate critical damping values are used for a specific component or subsystem, where documented test data justifies such usage. 3.7(B).4 Seismic Instrumentation In the event of an earthquake, the seismic instrumentation will provide information on the input ground motion and resultant vibratory responses of representative Category I structures and equipment so that an evaluation can be made as to: Whether input design response spectra were exceeded Whether the vibratory responses of the representative Category I structures and equipment were exceeded The need for shutdown of the plant The degree of validity of the mathematical models used in the seismic analysis of the buildings and equipment. The design requirement of the seismic instrume ntation is based on a Safe Shutdown Earthquake (SSE) of 0.25g (peak) and time duration of fifteen (15) seconds, and an Operating Basis Earthquake (OBE) of 0.13g (peak). The SM instrumentation is not Class 1E qualified.

It is not designed for a design basis event environment, except the SSE. It is ANS Safety Class NNS and Seismic Category I. 3.7(B).4.1 Comparison with Regulatory Guide 1.12 The SM system instrumentation type, number and locations described follow the guidance provided in USNRC Regulatory Guide 1.12, revision 1 (RG 1.12), 1974, with exceptions as discussed below. These exceptions are consistent with draft Regulatory Guide DG-1016, "Nuclear Power Plant Instrumentation for Earthquakes." Triaxial peak accelerographs, recommended by RG 1.12, for attachment to reactor equipment and piping inside containment, and to Seismic Category I equipment or piping outside containment, are not installed as part of the SM system. All locations where response spectrum record ers are recommended by Re gulatory Guide 1.12 are monitored by triaxial time history accelerographs. Triaxial time history accelerograph XT-6701, linked to the SSA-3 digital recording system, provides the seismic response of the Containment foundation.

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Page 25 RG 1.12 recommends the installation of accelerometers on two elevations (not including the foundation) of the Containment building. Triaxial time history acceler ographs are installed on the Containment foundation at elevation (-)26', and the operating deck at (+)25'. RG 1.12 recommends the installation of accelerometers on two independent Seismic Category I foundations (not including the Containment foundation). One triaxial time history accelerograph is installed in the Service Water Pump House at elevation 22'. One triaxial time history accelerograph is installed on elevation 53' in the PAB, not on two elevations of two independent Seismic Category I buildings as recommended by RG 1.12. 3.7(B).4.2 Location and Description of Instrumentation The SM instrumentation described below is intended to provide timely and accurate technical information necessary for an informed assessment of the integrity of sa fety-related components, systems and structures immediately following an earthquake. Electric power for seismic instrumentation is provided from a Class 1E, 120V, uninterruptable power supply. a. Time History Accelerograph System SSA-3/SSP-1 Seismic instrumentation is provided, one each at the following locations: 1. At a free field position in the control room east air intake, on bedrock. 2. Between columns 16 and 17 in the containment building foundation at elevation (-) 26'-0". 3. Between columns 16 and 17 on the concrete operating floor in the Containment Building at elevation (+)25'-0". The triaxial acceleromete r (type FBA-3) has a frequency range from 0.0 to 50.0 Hz. Each triaxial accelerometer is conn ected to a digital recorder that is part of the SSA-3/SSP-1 system located in seismic control panel 1-SM-CP-58. The trigger method is a threshold exceedance type. Its threshold level is set at 0.01g to avoid actuation due to insignificant motion but to record a seismic disturbance which creates significantly lower ground accelerations than that of the OBE (0.13g).

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 26 The two seismic instrumentation packages located in the Containment building are oriented so that the axes of the accelerometers are pointing in the same direction and are aligned to the axis of the building. The Kinemetrics solid state accelerograph system, SSA-3/SSP-1, is a central recording, time history accelerograph and playback system. The SSA-3 digital triggering system continuously monitors the signals sent by remote acceleration sensors. When motion exceeds the preprogrammed threshold, the recorders turn on and begin sampling concurrently on all channels. The data is automatically recorded with information such as sensor location, event time and sensor serial number. When the system detriggers, the SSP-1 interrogates the recorders and downloads the event files into a dedicated PC for automatic OBE analysis. The total capacity of the SSA-3 is 25 minutes. b. Solid State Accelerographs, SSA-1 Two stand-alone solid state accelerographs, SSA-1, are installed in the following locations: 1. In the Primary Auxiliary Building, PAB, at elevation 53'.

2. On the southwest corner of the electrical control room of the Service Water (SW) Pumphouse at elevation 22'-0". Each recorder can trigger and record a seismic event using a preset threshold level as described for the SSA-3 system. Each SSA-1 recorder has an internal triaxial accelerometer with a full scale of +2g, a frequency response from 0.0 to 50.0 Hz, and a 70% damping coefficient. c. Criteria for Instrument Location The selection of the above locations is based on the guidance provided in USNRC Regulatory Guide 1.12, Revision 1, for an SSE acceleration of less than 0.3g with exceptions as provided in Subsection 3.7(B).4.1. All instruments are accessible for inspection, test and service. Table 3.7(B)-24 summarizes the location of each primary instrument.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 27 3.7(B).4.3 Control Room Operator Notifications Audio visual alarms are provided on the main control board for the following parameters: a. "SEISMIC EVENT IN PROGRESS" b. "SEISMIC MONITOR LOSS OF AC/DC" These conditions apply to instruments at the fo llowing locations: 1. East air intake

2. Containment building foundation
3. Containment operating floor alarmed as "Seismic Event in Progress" 3.7(B).4.4 Comparison of Measured and Predicted Responses The SSP-1 is an automatic data retrieval and analysis system based on a high-speed, rack-mounted computer. The system remains on at all times monitoring the triggering activity of the SSA-3. When the SSA-3 has recorded an event, the SSP-1 retrieves the time history data and automatically evaluates OBE exceedance criteria including a computation of response spectra, cumulative absolute velocity (CAV) and spectral velocity for each axis (X,Y,Z). A graphics printer allows immediate print and plot results for operator review. Detailed comparisons are made between SM system measured responses and calculated responses based on plant dynamic models.

The time history records from the east air intake are used to ca lculate response spectra at the appropriate critical damping ratio. This is compared to the design response spectra.

The response spectra measured at the Service Water Pumphouse foundation and the Primary Auxiliary Building at elevation 53' are compared with those calculated by using the time history records from the east air intake as input ground motion to the containment and primary auxiliary building dynamic models. These comparisons indicate the validity of the dynamic model and form the basis for adjustment of the model.

Measured structural responses and response spectra are compar ed against the original design and analysis parameters to permit evaluation of the seismic effect on structures and equipment. These comparisons provide the basis for a deta iled physical inspection of structures and equipment.

S EABROOK STATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Seismic Design Revision 11 Section 3.7(B)

Page 28 3.7(B).5 In-Service Surveillance Calibration and alignment on three orthoginal axes are performed prior to fuel loading to assure proper operation. Periodic testing and calibration is performed in accordance with Technical Requirements.

3.7(B).6 References

1. "Nuclear Reactors and Earthquakes," TI D-7024, Prepared by Lockheed Aircraft Corporation and Holmes & Narver, Inc., for the Division of Reactor Development, U. S. Atomic Energy Commission, Washington, D. C., August 1963. 2. Whitman, R.V., 1969: "Equivalent Lumped System for Structure Founded Upon Stratum of Soil," Proc. 4th World C onf. on Earthquake Engineering, Chile, Vol. III, Section A-6, pp. 133-142. 3. Newmark, N. M. and Rosenblueth, E., "Fundamentals of Earthquake Engineering," Prentice-Hall, 1971. 4. Shah, H. H. and Chu, S. L., "Seism ic Analysis of Underground Structural Elements," Journal of the Power Division, Proceedings of the American Society of Civil Engineers, Vol. 100, No. P01, July 1974, PP. 53-62. 5. Iqbal, M. A. and Goodling, E. C., "Seismic Design of Buried Piping," and ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, December 1975.

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Page 1 3.7(N) SEISMIC DESIGN In addition to the steady-state loads imposed on the system under normal operating conditions, the design of equipment and equipment supports re quires that consideration also be given to abnormal loading conditions such as earthquakes. Seismic loadings are considered for earthquakes of two magnitudes: Safe Shut down Earthquake (SSE) and Operating Basis Earthquake (OBE). The SSE is defined as the maximum vibratory ground motion at the plant site that can reasonably be predicted from geologic and seismic evidence. The OBE is that earthquake which, considering the local geology and seismology, can be reasonably expected to occur during the plant life. For the OBE loading condition, the Nuclear Steam Supply System is designed to be capable of continued safe operation. The design for the SSE is intended to assure: a. That the integrity of the reactor coolant pressure boundary is not compromised

b. That the capability to shutdown the reactor and maintain it in a safe condition is not compromised c. That the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100 is not compromised. The seismic qualification requirements for safety-related instrumentation and electrical equipment are covered in Section 3.10. The seismic and ANS safety class definitions and classification lists are given in Section 3.2. Operating condition categories and the methods used for seismic qualification of mechanical equipment are given in Section 3.9.

3.7(N).1 Seismic Input 3.7(N).1.1 Design Response Spectra Refer to Subsection 3.7(B).1.1.

3.7(N).1.2 Design Time History Refer to Subsection 3.7(B).1.2.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 2 3.7(N).1.3 Critical Damping Values The damping values given in Table 3.7(N)-1 were used for the systems analysis of Westinghouse equipment. These are consistent with the damping values recommended in Regulatory Guide 1.61, except in the case of the primary coolant loop system components and large piping (excluding reactor pressure vessel internals) for which the damping values of 2 percent and 4 percent are used, as established in testing programs reported in Reference 1. (The damping values of ASME B&PV Code, Code Case N-411, ma y be used for pipe stress verification and for pipe support optimization in place of the foregoing). The damping values for control rod drive mechanisms (CRDM) and the fuel assemblies of the Nuclear Steam Supply System, when used in seismic system analysis, are in conformance with the values for welded and/or bolted steel structures (as appropriate) listed in Regulatory Guide 1.61. Tests on fuel assembly bundles justified conservative component damping values. Documentation of the fuel assembly tests is found in Reference 7.

The damping values used in component analysis of CRDM and their seismic supports were developed by testing programs performed by Westinghouse. These tests were performed during the design of the CRDM support; the support was designed so that the damping in Table 3.7(N)-1 could be conservatively used in the seismic analysis. The CRDM support system is designed with plates at the top of the mechanism and gaps between mechanisms. These are encircled by a box section frame which is attached by tie-rods to the refueling cavity wall. The test conducted was on a full-size CRDM complete with rod position indicator coils, attachment to a simulated vessel head, and variable gap between the top of the pressure housing support plate and a rigid bumper representing the support.

The internal pressure of the CRDM was 2250 psi and the temperature on the outside of the pressure housing was 400 F. The program consisted of transient vibration te sts in which the CRDM was deflected a specified initial amount and suddenly released. A logarithmic decrement analysis of the decaying transient provides the effective damping of the assembly. The effect on damp ing of variations in the drive shaft axial position, upper seismic support clearance and initial deflection amplitude was investigated. The upper support clearance had the largest effect on the CRDM damping with the damping increasing with increasing clearance. With an upper clearance of 0.06 inches, the measured damping was approximately 8 percent. The clearances in a typical upper seismic CRDM support is a minimum of 0.10 inches. The increasing damping with increasing clearances trend from the test results indicated that the damping would be greater than 8 percent for both the OBE and the SSE based on a comparison between typical deflections during these seismic events to the initial deflections of the mechanisms in the test. Component damping values of 5 percent are, therefore, conservative for both OBE and SSE.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 3 These damping values are used and applied to CRDM component analyses by response spectra techniques.

3.7(N).1.4 Supporting Media for Seismic Category I Structures Refer to Subsection 3.7(B).1.4.

3.7(N).2 Seismic System Analysis This section describes the methods of seismic analysis performed for safety-related components and systems within Westinghouse's scope.

3.7(N).2.1 Seismic Analysis Methods Those components and systems that must remain functional in the event of the Safe Shutdown Earthquake (Seismic Category I) are identified by applying the criteria of Subsection 3.2.1. In general, the dynamic analyses are performed using a modal analysis plus either the response spectrum analysis or integration of the uncoupled modal equations as described in Subsections 3.7(N).2.1c and 3.7(N).2.1d, respectively, or by direct integration of the coupled differential equations of motion described in Subsection 3.7(N).2.1e.

a. Dynamic Analysis - Mathematical Model The first step in any dynamic analysis is to model the structure or component, i.e., convert the real structure or component into a system of masses, springs, and dash pots suitable for mathematical analysis. Th e essence of this step is to select a model so that the displacements obtai ned will be a good representation of the motion of the structure or component. Stat ed differently, the true inertia forces should not be altered so as to appreciably affect the internal stresses in the structure or component. Some typical modeling techniques are presented in

Reference 3.

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UFSAR Seismic Design Page 4 iiri i i rirr0ukucÿm 1. Equations of Motion Consider the multi-degree of freedom system shown in Figure 3.7(N)-1. Making a force balance on each mass point r, the equations of motion can be written in the form:

(3.7(N)-1) m r = the value of the mass or mass moment of rotational inertia at mass point r

ÿr = absolute translational or angular acceleration of mass point r c ri = damping coefficient - external force or moment required at mass point r to produce a unit translational or angular velocity at mass point i, maintaining zero translational or angular velocity at all other mass points. Force or moment is positive in the direction of positive translational or angular velocity.

u i = translational or angular velocity of mass point i relative to the base k ri = stiffness coefficient - the external force (moment) required at mass point r to produce a unit deflection (rotation) at mass point i, maintaining zero displacement (rotation) at all other mass points. Force (moment) is positive in the direction of positive displacement (rotation).

u i = displacement (rotation) of mass point i relative to the base. As an example, note that Figure 3.7(N)-1 does not attempt to show all of the springs (and none of the dashpots) which are represented in

Equation (3.7(N)-1).

Since: ÿr = ür + ÿs (3.7(N)-2) Where: ÿs = absolute translational (angular) acceleration of the base S EABROOK D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND Revision 8 SYSTEMS S TATION Section 3.7(N)

UFSAR Seismic Design Page 5 ÿm u k ucüm r i i ri i i ri r r su ÿr = translational (angular) acceleration of mass point r relative to the base Equation (3.7(N)-1) can be written as:

(3.7(N)-3) For a single degree of freedom system with displacement u, mass m, damping c, and stiffness k, the corresponding equation of motion is: mü +c+ku = -mÿs (3.7(N)-4)

b. Modal Analysis
1. Natural Frequencies and Mode Shapes The first step in the modal analysis method is to establish the normal modes, which were determined by eigen solution of Equation (3.7(N)-3). The right hand side and the damping term are set equal to zero for this purpose as illustrated in Reference 4 (pp. 83 through 111). Thus, Equation (3.7(N)-3) becomes: 0ukümiri irr0KM

(3.7(N)-5) The equation given for each mass point r in Equation (3.7(N)-5) can be written as a system of equations in matrix form as:

(3.7(N)-6) where: [M] = mass and rotational inertia matrix

{} = column matrix of the general displacement and rotation at each mass point relative to the base D ESIGN OF S EABROOK S TATION UFSAR STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 6 [K] = square stiffness matrix

{ } = column matrix of general translational and angular accelerations at each mass point relative to the base, d 2 {} dt 2. Harmonic motion is assumed and the {} is expressed as:

{} = {} sin t (3.7(N)-7) where: {} = column matrix of the spatial displacement and rotation at each mass point relative to the base

= natural frequency of harmonic motion in radians per second The displacement function and its second derivative are substituted into Equation (3.7(N)-6) and yield:

[K] {} = 2 [M] {} (3.7(N)-8) The determinant [K] - 2 [M] is set equal to zero and is then solved for the natural frequencies. The associated mode shapes are then obtained from Equation (3.7(N)-8). This yields n natural frequencies and mode shapes where n equals the number of dynamic degrees of freedom of the system. The mode shapes are all orthogonal to each other and are sometimes referred to as normal mode vibrations. For a single degree of freedom system, the stiffness matrix and mass matrix are single terms and the determinant [K] - 2 [M] when set equal to zero yields simply:

k - 2m = 0 or:

(3.7(N)-9) m k where is the natural angular frequency in radians per second. The natural frequency in cycles per second is therefore:

m k 21 f (3.7(N)-10)

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UFSAR Seismic Design Page 7 To find the mode shapes, the natural frequency corresponding to a particular mode, n , can be substituted in Equation (3.7(N)-8). 2. Modal Equations The response of a structure or component is always some combination of its normal modes. Good accuracy can usually be obtained by using only the first few modes of vibration. In the normal mode method, the mode shapes are used as principal coordinates to reduce the equations of motion

to a set of uncoupled differential equations that describe the motion of each mode n. These equations may be written as (Reference 4, pp. 116 125):

n + 2n p n A + n 2 A n = - n ÿs (3.7(N)-11) where the modal displacement or rotation, A n , is related to the displacement or rotation of mass point r in mode n, u rn , by the equation:

u rn + A n rn (3.7(N)-12) where: n = natural frequency of mode n in radians per second.

P n = critical damping ratio of mode n.

n = modal participation factor of mode n given by n rn 2 r m'rnr m n n (3.7(N)-13) where:

_rn = value of rn in the direction of the earthquake. The essence of the modal analysis lies in the fact that Equation (3.7(N)-11) is analogous to the equation of motion for a single degree of freedom system that will be developed from Equation (3.7(N)-4). Dividing Equation (3.7(N)-4) by m gives:

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UFSAR Seismic Design Page 8 sÿu m k u m cü (3.7(N)-14) The critical damping ratio of the single degree of freedom system, p, is defined by the equation:

c c c p (3.7(N)-15) where the critical damping coefficient is given by the expression:

m2c c (3.7(N)-16) Substituting Equation (3.7(N)-16) into Equation (3.7(N)-15) and solving for c/m gives: p2 cm s 2ÿup2ü (3.7(N)-17) Substituting this expression and the expression for k/m given by Equation (3.7(N)-9) into Equation (3.7(N)-14) gives:

(3.7(N)-18) Note the similarity of Equations (3.7(N)-11) and (3.7(N)-18). Thus, each mode may be analyzed as though it were a single degree of freedom system and all modes are independent of each other. By this method a fraction of critical damping, i.e., c/c c, may be assigned to each mode and it is not necessary to identify or evaluate individual damping coefficients, i.e., c. However, assigning only a single damping ratio to each mode has a drawback. There are three ways used to overcome this limitation when considering a slightly damped st ructure (e.g., steel) supported by a massive moderately damped structure (e.g., concrete).

D ESIGN OF STRUC S EABROOK S TATION UFSAR TURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 9 The first method is to develop and analyze separate mathematical models for both structures using their respective damping values. The massive moderately damped support structure is analyzed first. The calculated response at the support points for the slightly damped structures is used as a forcing function for the subsequent detailed analysis. The second method is to inspect the mode shapes to determine which modes correspond to the slightly damped structure and then use the damping

associated with the structure having predominant motion. The third method is to use the Rayleigh damping method based on computed modal energy distribution.

c. Response Spectrum Analysis The response spectrum is a plot showing the variation in the maximum response (Reference 4, pages 24 through 51) (displacem ent, velocity, and acceleration) of a single degree of freedom system versus it s natural frequency of vibration when subjected to a time history motion of its base. The response spectrum concept can be best explained by outlining the steps involved in developing a spectrum curve. Determination of a single point on the

curve requires that the response (displacem ent, velocity, and acceleration) of a single degree of freedom system with a given damping and natural frequency is calculated for a given base motion. The variations in response are established and the maximum absolute value of each is pl otted as an ordinate with the natural

frequency used as the abscissa. The process is repeated for other assumed values of frequency in sufficient detail to establish the complete curve. Other curves corresponding to different fractions of critical damping are obtained in a similar fashion. Thus, the determination of each point of the curve requires a complete dynamic response analysis, and the determination of a complete spectrum may involve hundreds of such analyses. However, once a response spectrum plot is generated for the particular base motion, it may be used to analyze each structure and component with the base motion.

The spectral acceleration, velocity, and displacement are related by the equation:

n d 2 n nvn n a SSS (3.7(N)-19)

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UFSAR Seismic Design Page 10 There are two types of response spectra that must be considered. If a given building is shown to be rigid and to have a hard foundation, the ground response spectrum or ground time history is used.

It is referred to as a ground response spectrum. If the building is flexible and/or has a soft foundation, the ground response spectrum is modified to include these effects. The response spectrum at various support points must be developed. These are called floor response

spectra. d. Integration of Modal Equations This method can be separated into the following two basic parts: Integration procedure for the uncoupled modal Equation (3.7(N)-11) to obtain the modal displacements and accelerations as a function of time. Using these modal displacements and accelerations to obtain the total displacements, accelerations, forces, and stresses. 1. Integration Procedure Integration of these uncoupled modal equations is done by step-by-step numerical integration. The step-by-step numerical integration procedure consists of selecting a suitable time interval, t, and calculating modal acceleration, n, modal velocity, n, and modal displacement, A n , at discrete time stations t apart, starting at t = 0 and continuing through the range of interest for a given time history of base acceleration. A2. Total Displacements, Accelerations, Forces and Stresses From the modal displacements and accelerations, the total displacements, accelerations, forces, and stresses can be determined as follows: (a) Displacement of mass point r in mode n as a function of time is given by Equation (3.7(N)-12) as:

u rn = A n rn (3.7(N)-20) with the corresponding acceleration of mass point r in mode n as:

u rn = A n rn (3.7(N)-21)

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UFSAR Seismic Design Page 11 (b) The displacement and acceleration values obtained for the various modes are superimposed algebraically to give the total displacement and acceleration at each time interval. (c) The total acceleration at each time interval is multiplied by the mass to give an equivalent static force. Stresses are calculated by applying these forces to the model or from the deflections at each time interval.

e. Integration of Coupled Equations of Motion The dynamic transient analysis is a tim e history solution of the response of a given structure to known forces and/or displacement forcing functions. The structure may include linear or nonlinear elements, gaps, interfaces, plastic elements, and viscous and Coulomb dampers. Nodal displacements, nodal forces, pressure, and/or temperatures may be c onsidered as forcing functions. Nodal displacements and elemental stresses for the complete structure are calculated as functions of time. The basic equations for the dynamic analysis are as follows:

[M] {} + [C] } + [K] {x} = {F(t)} xxxxz(3.7(N)-22) where the terms are as defined earlier and {F(t)} may include the effects of applied displacements, forces, pressures, temperatures, or nonlinear effects such as plasticity and dynamic elements with gaps. Options of translational accelerations input to a structural system and the inclusion of static deformation and/or pre-load may be considered in the nonlinear dynamic transient analysis.

The option of translational input such as uniform base motion to a structural system is considered by introducing an inertia force term of -[M]{z} to the right

hand side of the basic Equation (3.7(N)-22), i.e., [M] { } = [C] {

} + [K] {x} = {F} - [M] {} (3.7(N)-23)

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UFSAR Seismic Design Page 12 The vector {z} is defined by its components z i where i refers to each degree of freedom of the system. z i is equal to a 1 , a 2 , or a 3 if the i-th degree of freedom is aligned with the direction of the system translational acceleration a 1 , a 2 , or a 3 , respectively. z i = 0 if the i-th degree of freedom is not aligned with any direction of the system translational acceleration.

Typical application of this option is a structural system subjected to a seismic excitation of a given ground acceleration record. The displacement {x} obtained from the solution of Equation (3.7(N)-23) is the displacement relative to the ground. The option of the inclusion of initial static deformation or preload in a nonlinear transient dynamic structural analysis is considered by solving the static problem prior to the dynamic analysis. At each stat e of integration in transient analysis, the portion of internal forces due to static deformation is always balanced by the portion of the forces which are statically applied. Hence, only the portion of the forces which deviate from the static loads will produce dynamic effects. The output of this analysis is the total result due to static and dynamic applied loads. One available method for the numerical in tegration of Equations (3.7(N)-22) and (3.7(N)-23) is the Newmark Beta integration scheme proposed by Chan, Cox, and Benfield (Reference 6). In this integration scheme, Equations (3.7(N)-22) and (3.7(N)-23) are replaced by: n2nn1n2n 2xx21xKCxxt2 1x2xxMt 1n1n2nn1n2nxF21F (3.7(N)-24) where: n, n+1, n+2 = past, present, and future (updated) values of the variables

= parameter to be selected on the basis of numerical stability and accuracy F = the total right hand side of the equation of motion (Equation (3.7(N)-22) or (3.7(N)-23))

t = t n+2 - t n+1 = t n+1 - t n S EABROOK D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND Revision 8 SYSTEMS S TATION Section 3.7(N)

UFSAR Seismic Design Page 13 The value of is chosen equal to 1/3 in order to provide a margin of numerical stability for nonlinear problems. Since the numerical stability of Equation (3.7(N)-24) is mostly determined by the left hand side terms of that equation, the right hand side terms were replaced by F n+2. Furthermore, since the time increment may vary between two successive time substeps, Equation (3.7(N)-24) may be modified as follows: 2nn1n2n 3 1n2n 1 1 1FxxxKxxCtt 1tt Mtt n1n1n2nxxxx 2 (3.7(N)-25) By factoring x n+2 , x n+1 , and x n, and rearranging terms, Equation (3.7(N)-25) is obtained as follows: 1n 3 1 7 n2n 3 3 5xKMC12FxKCCMC (3.7(N)-26) where: 1 1 2ttt 2 C 1 3tt 1 C 1 5ttt 2 C 527CCC S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 14 The above set of simultaneous linear equations is solved to obtain the present values of modal displacements {x t} in terms of the previous (known) values of the nodal displacements. Since [M], [C], and [K] are included in the equation, they can also be time or displacement dependent.

3.7(N).2.2 Natural Frequencies and Response Loads Refer to Subsection 3.7(B).2.2.

3.7(N).2.3 Procedures Used for Modeling Procedures used for modeling are discussed in Subsection 3.7(N).2.1a.

3.7(N).2.4 Soil/Structure Interaction Refer to Subsection 3.7(B).2.4.

3.7(N).2.5 Development of Floor Response Spectra For Westinghouse floor response spectra development scope, a seismic time history analysis using a coupled building-NSSS model is performed. The seismic input (one input for the OBE analysis and one input for the SSE analysis) is composed of three time history components which are applied simultaneously at the base mat location. The time history input is applied at the base mat location. Using the seismic response from the time history analysis, reactor coolant response spectra are developed as outlined in Subsection 3.7(N).2.1c. The reactor coolant system spectra is used to evaluate feedwater lines, main steam lines, and pressurizer safety and relief valve lines.

3.7(N).2.6 Three Components of Earthquake Motion The seismic design of the piping and equipment includes the effect of the seismic response of the supports, equipment, structures and components. The system and equipment response is determined using three earthquake components, two horizontal and one vertical. The design ground spectra, are the bases for generating these three input components. Floor response

spectra are generated for two perpendicular horizontal directions (i.e., N-S, E-W) and the vertical direction. System and equipment analysis is performed with these input components applied in the N-S, E-W and vertical directions. The damping values used in the analysis are those given in Table 3.7(N)-1. In computing the system and equipment response by response spectrum modal analysis, the methods of Subsection 3.7(N).2.7 are used to combine all significant modal responses to obtain the combined unidirectional responses.

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UFSAR Seismic Design Page 15 The combined total response is then calculated using the square-root-of-the-sum-of-the-squares formula applied to the resultant unidirectional responses. For instance, for each item of interest such as displacement, force, stresses, etc., th e total response is obtained by applying the above described method. The mathematical expression for this method (with R as the item of interest) is: 2 1 3 2RR (3.7(N)-27) 1T C Twhere 2 1 i N1i T T 2RR (3.7(N)-28) where R C = total combined response at a point R T = value of combined response of direction T R Ti = absolute value of response for direction T, mode i N = total number of modes considered The subscripts can be reversed without changing the results of the combination.

For the case of closely spaced modes, R T in Equation (3.7(N)-28) shall be replaced with R T as given by Equation (3.7(N)-29) in Subsection 3.7(N).2.7.

The system and equipment response can also be determined using time history analyses.

If the time history analysis is performed by appl ying the two horizontal and vertical time history components independently, the total combined response is computed adding algebraically the unidirectional responses at each time step. When the three time history components are applied simultaneously, the total response is obtained by direct time integration of the equations of motion.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 16 3.7(N).2.7 Combination of Modal Responses The total unidirectional seismic response is obtained by combining the individual modal responses utilizing the square-root-of-the-sum-of-the-squares method. For systems having modes with closely spaced frequencies, this method is modified to include the possible effect of these modes. The groups of closely spaced mode s are chosen such that the difference between the frequencies of the first mode and the last mode in the group does not exceed 10 percent of the lower frequency. Groups are formed starting from the lowest frequency and working towards

successively higher frequencies. No one frequency is in more than one group. Combined total response for systems which have such closely spaced modal frequencies is obtained by adding to the square root of the sum of the squares of all modes the product of the responses of the modes in each group of closely spaced modes and a coupling factor . This can be represented mathematically as: KK1K j jMK1j1j 2 i1i 2 TRRRRNNS N (3.7(N)-29) where: R T = total unidirectional response R i = absolute value of response of mode i N = total number of modes considered S = number of groups of closely spaced modes

M j = lowest modal number associated with group j of closely spaced modes N j = highest modal number associated with group j of closely spaced modes K = coupling factor with 12KK KK1, (3.7(N)-30)

D ESIGN OF STRUCTURES , C OMPONENTS , E S EABROOK S TATION UFSAR QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 17 and: 2 1 2 KK'1K' (3.7(N)-31) d K 2'tKK6776344324422332 3 8 2 3 2 2 2 1 2 TRR2RR2RR2RRRRRRR (3.7(N)-32) where: K = frequency of closely spaced mode K K = fraction of critical damping in closely spaced mode K t d = duration of the earthquake An example of this equation applied to a system can be supplied with the following considerations. Assume that the predominant contributing modes have frequencies as given below: Mode 1 2 3 4 5 6 7 8 Frequency 5.0 8.0 8.3 8.6 11.0 15.5 16.0 20.0 There are two groups of closely spaced modes, namely with modes (2, 3, 4) and (6, 7).

Therefore: S = 2, number of groups of closely spaced modes M 1 = 2, lowest modal number associated with group 1 N 1 = 4, highest modal number associated with group 1 M 2 = 6, lowest modal number associated with group 2 N 2 = 7, highest modal number associated with group 2 N = 8, total number of modes considered The total response for this system is, as derived from the expansion of Equation (3.7(N)-29):

(3.7(N)-33)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 18 3.7(N).2.8 Interaction of Non-Category I Structures with Seismic Category I Structures Refer to Subsection 3.7(B).2.8.

3.7(N).2.9 Effects of Parameter Variations on Floor Response Spectra Refer to Subsection 3.7(B).2.9.

3.7(N).2.10 Use of Constant Vertical Static Factors Constant vertical static factors are not used on the vertical floor response load for the seismic design of safety classed systems and components within Westinghouse's scope of responsibility. All such systems and components are analyzed in the vertical direction.

3.7(N).2.11 Methods Used to Account for Torsional Effects Refer to Subsection 3.7(B).2.11.

3.7(N).2.12 Comparison of Responses Refer to Subsection 3.7(B).2.12.

3.7(N).2.13 Methods for Seismic Analysis of Dams Refer to Subsection 3.7(B).2.13.

3.7(N).2.14 Determination of Seismic Category I Structure Overturning Moments Refer to Subsection 3.7(B).2.14.

3.7(N).2.15 Analysis Procedure for Damping In instances under the standard scope of Wes tinghouse supply and analysis, either the lowest damping value associated with the elements of the system is used for all modes, or an equivalent modal damping value is determined by testing programs such as was done for the reactor coolant loop (Reference 1). Procedures for damping are further discussed in Subsection 3.7(N).2.1.

3.7(N).3 Seismic Subsystem Analysis This section describes the seismic analysis performed on subsystems within Westinghouse's scope of responsibility.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 19 3.7(N).3.1 Seismic Analysis Methods Seismic analysis methods for subsystems within Westinghouse scope of responsibility are given in Subsection 3.7(N).2.1.

3.7(N).3.2 Determination of the Number of Earthquake Cycles For each OBE, the system and components will have a maximum response corresponding to the maximum induced stresses. The effect of these maximum stresses for the total number of OBEs must be evaluated to assure resistance to cyclic loading. The OBE is conservatively assumed to occur 20 times over the life of the plant. The number of maximum stress cycles for each occurrence depends on the system and component damping values, complexity of the system and component, duration and frequency contents of the input earthquake. A precise determination of the number of maximum stress cycles can only be made using time history analysis for each item, which is not feasible. Instead, a time history study has been conducted to arrive at a realistic number of maximum stress cycles for all Westinghouse systems and components. To determine the conservative equivalent number of cycles of maximum stress associated with each occurrence, an evaluation was performed considering both equipment and its supporting building structure as single-degree-of-freedom systems. The natural frequencies of the building and the equipment are conservatively chosen to coincide. The damping in the equipment and building are equivalent to the damping values in Table 3.7(N)-1.

The results of this study indicate that the total number of maximum stress cycles in the equipment having peak acceleration above 90 percent of the maximum absolute acceleration did not exceed eight cycles. If the equipment was assumed to be rigid in a flexible building, the number of cycles exceeding 90 percent of the maximum stress was not greater than three cycles.

This study was conservative since it was performed with single-degree-of-freedom models which tend to produce a more uniform and unattenuated response than a complex interacting system. The conclusions indicate that 10 maximum stress cycles for flexible equipment (natural frequencies less than 33 Hz.) and 5 maximum stress cycles for rigid equipment (natural frequencies greater than 33 Hz.) for each of 20 OBE occurrences should be used for fatigue evaluation of Westinghouse systems and components.

3.7(N).3.3 Procedure Used for Modeling Refer to Subsection 3.7(N).2.1a for modeling procedures for subsystems in Westinghouse scope of responsibility.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 20 3.7(N).3.4 Basis for Selection of Frequencies The analysis of equipment subjected to seismic lo ading involves several basic steps, the first of which is the establishment of the intensity of the seismic loading. Considering that the seismic input originates at the point of support, the response of the equipment and its associated supports based upon the mass and stiffness characteristics of the system, will determine the seismic accelerations which the equipment must withstand. Three ranges of equipment/support behavior which affect the magnitude of the seismic acceleration are possible: a. If the equipment is rigid relative to the structure, the maximum acceleration of the equipment mass approaches that of the structure at the point of equipment support. The equipment acceleration value in this case corresponds to the low-period region of the floor response spectra. b. If the equipment is very flexible relativ e to the structure, the equipment will show very little response. c. If the periods of the equipment and supporting structure are nearly equal, resonance occurs and must be taken into account. In all cases, equipment under earthquake loadings is designed to be within code allowable stresses. Also, as noted in Subsection 3.7(N).3.2, rigid equipment/support systems have natural frequencies greater than 33 Hz.

3.7(N).3.5 Use of Equivalent Static Load Method of Analysis The static load equivalent or static analysis method involves the multiplication of the total weight of the equipment or component member by the specified seismic acceleration coefficient. The magnitude of the seismic acceleration coeffici ent is established on the basis of the expected dynamic response characteristics of the component. Components which can be adequately characterized as single-degree-of-freedom systems are considered to have a modal participation factor of one. Seismic acceleration coefficients for multi-degree of freedom systems which may be in the resonance region of the amplified response spectra curves are increased by 50 percent to account conservatively for the increased moda l participation and the contributions of the higher modes of vibration.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 21 3.7(N).3.6 Three Components of Earthquake Motion Methods used to account for three components of earthquake motion for subsystems in Westinghouse's scope of responsibility are given in Subsection 3.7(N).2.6.

3.7(N).3.7 Combination of Modal Responses Methods used to combine modal responses for subsystems in Westinghouse's scope of responsibility are given in Subsection 3.7(N).2.7.

3.7(N).3.8 Analytical Procedures for Piping The Class I piping systems are analyzed to the rules of the ASME Code,Section III, NB 3650. When response spectrum methods are used to evaluate piping systems supported at different elevations, the following procedures are used. The effect of differential seismic movement of the piping supports is included in the piping anal ysis according to the rules of the ASME Boiler and Pressure Vessel Code,Section III, Paragra ph NB 3653. According to ASME definitions, these displacements cause secondary stresses in the piping system. The response quantity of interest induced by differential seismic motion of the support is computed statically by considering the building response on a mode-by-mode basis. In the response spectrum dynamic analysis for evaluation of piping systems supported at different elevations, spectra which envelop the floor response spectra corresponding to the applicable support locations or nozzles are used. Westinghouse does not have in their scope of analysis any piping systems interconnected between buildings.

3.7(N).3.9 Multiply-Supported Equipment Components with Distinct Inputs When response spectrum methods are used to evaluate reactor coolant system primary components interconnected between floors, the procedure of the following paragraphs is used. The primary components of the Reactor Coolant System are supported at no more than two floor elevations. A dynamic response spectrum analysis is first made assuming no relative displacement between support points. The response spectra used in this analysis is the most severe floor response spectra.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 22 Secondly, the effect of differential seismic movement of components interconnected between floors is considered statically in the integrated system analysis and in the detailed component analysis. The results of the building analysis are reviewed on a mode-by-mode basis to determine the differential motion in each mode.

Per ASME Code rules, the stress caused by differential seismic motion is clearly secondary for piping (NB 3650) and component supports (NF 3231). For components, the differential motion will be evaluated as a free end displacement, since, per NB 3213.19, examples of a free end displacement are motions "that would occur because of relative thermal expansion of piping, equipment, and equipment supports, or because of rotations imposed upon the equipment by sources other than the piping." The effect of the differential motion is to impose a rotation on the component from the building. This motion, then, being a free end displacement and being similar to thermal expansion loads, will cause stresses which will be evaluated with ASME Code methods including the rules of NB 3227.5 used for stresses originating from restrained free end displacements. The results of these two steps, the dynamic inertia analysis and the static differential motion analysis, are combined absolutely with due c onsideration for the ASME classification of the stresses.

3.7(N).3.10 Use of Constant Vertical Static Factors Constant vertical load factors are not used on the vertical floor response load for the seismic design of safety-related components and equipment within Westinghouse's scope of responsibility.

3.7(N).3.11 Torsional Effects of Eccentric Masses The effect of eccentric masses, such as valves and valve operators, is considered in the seismic piping analyses. These eccentric masses are modeled in the system analysis and the torsional effects caused by them are evaluated and included in the total system response. The total response must meet the limits of the criteria applicable to the safety class of the piping.

3.7(N).3.12 Buried Seismic Category I Piping Systems and Tunnels Refer to Subsection 3.7(B).3.12.

3.7(N).3.13 Interaction of Other Piping with Seismic Category I Piping Refer to Subsection 3.7(B).3.13.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 23 3.7(N).3.14 Seismic Analysis for Reactor Internals Fuel assembly component stresses induced by horizontal seismic disturbances are analyzed through the use of finite element computer modeling. The time history floor response based on a standard seismic time history normalized to SSE levels is used as the seismic input. The reactor internals and the fuel assemblies are modeled as spring and lumped mass systems or beam elements. The component seismic response of the fuel assemblies is analyzed to determine design adequacy. A detailed discussion of the analyses performed for typical fuel assemblies is contained in Reference 7. Fuel assembly lateral structural damping obtained experimentally is presented in Reference 7 (Figure B-4). The distribution of fuel assembly amplitudes decreases as one approaches the center of the core. Fuel assembly displacement time history for the SSE seismic input is illustrated in Reference 7 (Figure 2-3).

The CRDMs are seismically analyzed to confirm that system stresses under the combined loading conditions, as described in Subsection 3.9(N).1, do not exceed allowable levels, as defined by the ASME Code,Section III for "Ups et" and "Faulted" conditions. The CRDM is mathematically modeled as a system of lumped and distributed masses. The model is analyzed under appropriate seismic excitation, and the resultant seismic bending moments along the length of the CRDM are calculated. The corresponding stresses are then combined with the stresses from the other loadings required, and the combination is shown to meet the ASME Code,Section III, requirements.

3.7(N).3.15 Analysis Procedure for Damping Analysis procedures for damping for subsystems in Westinghouse's scope of responsibility are

given in Subsection 3.7(N).2.15.

3.7(N).4 Seismic Instrumentation Refer to Subsection 3.7(B).4.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Design Revision 8 Section 3.7(N)

Page 24 3.7(N).5 References

1. "Damping Values of Nuclear Plant Components," WCAP-7921-AR, May 1974. 2. Geninski, T. L. and Chiang, D., "Safety Analysis of the 17x17 Fuel Assembly for combined Seismic and Loss-of-Coolant Accident," WCAP-8236 (Proprietary), December 1973 and WCAP-8286 (Nonproprietary), December 1974. 3. Lin, C. W., "How to Lump the Masses - A Guide to the Piping Seismic Analysis," ASME paper 74-NE-7 presented at the Pr essure Vessels and Piping Conference, Miami, Florida, June 1974. 4. Biggs, J. M., Introduction to Structural Dynamics , McGraw-Hill, New York, 1964. 5. Thomas, T. H., et al., "Nuclear Reactors and Earthquakes," TID-7024, U.S. Atomic Energy Commission, Washington, D. C., August 1963. 6. Chan, S. P., Cox, H. L. and Benfield, W. A., "Transient Analysis of Forced Vibration of Complex Structural-Mechanical Systems," J. Royal Aeronautical Society," July 1962. 7. "Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March 1974 and WCAP-8288, Addendum 1 (Non-proprietary), April 1974.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 1 3.8 DESIGN OF CATEGORY I STRUCTURES

3.8.1 Concrete

Containment The containment structure houses the major portion of a PWR Nuclear Steam Supply System (NSSS). During the operating life of the plant, it will also provide the following functions:

a. Limiting the leakage rate to the maximum allowable Type "A" test leakage rate, 0.15 percent by weight of the containment contained air mass per day at calculated peak pressure and associated temperature, resulting from any loss-of-coolant accident (LOCA) and other postulated accidents.
b. Providing continuing radiation shielding during normal plant operation in accordance with 10 CFR 20 and during accident conditions in accordance with 10 CFR 100. c. Protecting the reactor vessel and all other safety

-related systems, equipment and components located inside the containment against all postulated external environmental conditions and resulting loads.

3.8.1.1 Description of Containment The containment, Figure 1.2-2, Figure 1.2-3, Figure 1.2-4, Figure 1.2-5 and Figure 1.2-6, is a seismic Category I reinforced concrete dry structure, which is designed to function at atmospheric conditions. It consists of an upright cylinder topped with a hemispherical dome, supported on a reinforced concrete foundation mat which is keyed into the bedrock by the depression for the reactor pit and by continuous bearing around the periphery of the foundation mat. The inside diameter of the cylinder is 140 feet and the inside height from the top of the base mat to the apex of the dome is approximately 219 feet; the net free volume is approximately 2,704,000 cubic feet. A welded steel liner plate, anchored to the inside face of the containment, serves as a leaktight membrane. Although not a code requirement, welds that are embedded in concrete and not readily accessible are covered by a leak chase system which permits leak testing of those welds throughout the life of the plant. Exemptions to these inaccessible welds are the welds joining mechanical penetrations X

-60 and X-61 to the steel liner plate. (The venting pipes which join the leak chase channels for these penetrations to the atmosphere were not provided; however, these welds underwent proper testing before they became inaccessible.) The liner on top of the foundation mat is protected by a four feet thick concrete fill mat which supports the containment internals and forms the floor of the containment.

The containment is designed to assure that the base mat, cylinder, and dome behave integrally to resist all loads.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 2 Located outside the Containment Building and having a similar geometry is the Containment Enclosure Building. This structure provides leak protection for the containment and protects it from certain loads, as discussed in Subsection 3.8.1.3. The Containment Enclosure Building is described in Subsection 3.8.4. a. Base Mat The reinforced concrete base mat is 153 feet in diameter and 10 feet thick. It is designed to carry the loads from the shell of the containment and from the internal structures.

An orthogonal grid rebar arrangement is provided for the bottom face of the base mat to simplify fabrication and construction. A radial and hoop pattern is used at the top face to minimize interference with cylinder dowels. Vertical and inclined shear reinforcement are provided to resist the transverse shear forces caused by design accident pressure and seismic loads. Details of the base mat reinforcing steel are shown on Figure 3.8-1. The mat liner plate is 1/4" thick with joints welded to leveling angles which serve as welding backup strips.

Internal structures are supported on and anchored to the fill mat, as indicated above. The mat is not anchored to the base mat. Stability of the containment with internals is provided by the keying action of the base mat and reactor pit in the rock and by bearing against the foundation for the Enclosure Building, which in turn transfers all horizontal shears directly into the bedrock through fill concrete.

b. Cylinder The cylinder has an inside diameter of 140 feet and is nominally 4'

-6" thick. Also, it is thickened to provide room for additional reinforcing steel around the openings for the equipment hatch and the personnel air lock.

The reinforcing bars in the cylinder are arranged and oriented to resist hoop, meridional and shear forces, including hoop, meridional and radial shear forces produced by bending moments. Orthogonal layers of bars in the meridional and hoop directions are provided on each face to resist the membrane forces primarily from pressure and seismic loads.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 3 An orthogonal set of bars inclined at 45 degrees to the horizontal is provided on the outside face to resist in

-plane seismic shear forces and membrane tension from other loads. Near the base of the containment, additional meridional bars and radial inclined stirrups are provided to resist discontinuity moments and radial shears, respectively, caused by the restraint on the cylinder at the junction of the cylinder and the base mat. Stirrups are also provided at the springline to resist radial shear.

Where there are large openings for access hatchways and penetrations, the main reinforcing bars are continued without interruption around the openings. No main reinforcing bars are terminated at any opening. Furthermore, additional bars are provided to resist the local effects of these openings and, around large openings such as the equipment hatch (28'

-0" inside diameter) and personnel air lock (7'-11/4" inside diameter), the concrete is thickened locally to resist the additional local forces and to accommodate the additional reinforcing.

Basic details of reinforcing steel in the cylinder are shown on Figure 3.8-2, which also includes reinforcing steel in the dome and junction of the wall and base mat. Details of the reinforcing steel at the equipment hatch and personnel air lock are shown on Figure 3.8-3 and Figure 3.8-4, respectively. These figures also show the transition of reinforcing steel between the openings and the membrane regions. The liner plate in the cylinder is 3/8" thick in all areas except penetrations and the junction of the base mat and cylinder where it is 3/4" thick. The liner is provided with an anchorage system to assure that it can withstand accident loadings while maintaining leak tightness. In addition, the anchorage system assures that the liner, which is used as a form during construction, can resist the hydrostatic concrete loads while maintaining liner tolerances within allowable values. The anchorage system consists of vertical tees spaced every 20 inches around the circumference of the cylindrical wall. The webs of the tees are welded to the liner plate with two 1/4" continuous fillet welds. Bent studs are attached to the flange of vertical tees as required to accommodate placement of rebar. Liner details are shown on Figure 3.8-5.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 4 Containment penetrations, other than the equipment hatch and personnel air lock, are located in the lower portion of the cylindrical structure. In general, a penetration consists of a sleeve anchored in the concrete cylinder wall and welded to the locally thickened containment liner. The weld between the liner and the sleeve is covered by a leak chase system which can be pressurized to demonstrate the integrity of the penetration

-to-liner weld joint. The piping, electrical cable and instrumentation cable pass through the embedded sleeves and the ends of the resulting annuli are closed off either by welded end plates or a flued head welded to the sleeve outside the containment. If the pipe carries hot fluid, the space between the pipe and the sleeve is insulated to maintain the concrete temperature adjoining the embedded sleeve at or below 200 F during normal plant operation. The fuel transfer tube passes through an embedded sleeve which has its ends closed off by an expansion bellows and an end plate. In the case of ventilation ducts, the sleeve forms the wall of the duct.

Sleeves for all penetrations, including the equipment hatch and personnel air lock, are embedded in the concrete wall by an engineered anchorage system that is welded to the penetration sleeve. Reinforcing steel, hoop, meridional and diagonal, is splayed around penetrations permitting all bars to be continuous. See Figure 3.8-6 for details of reinforcing at penetrations.

All brackets and attachments are welded to attachment plates which are welded to the liner plate and anchored into the concrete by studs welded on the opposite side of the liner, thereby transmitting the forces directly to the concrete. See Figure 3.8-5. In the reactor cavity pit, brackets are used, as a construction aid, on the concrete side of the liner to temporarily support the reinforcing steel. This is a permanent attachment which only functions until the concrete is placed. Each of these brackets is welded to an attachment plate; the attachment plate is welded to the liner with a continuous fillet weld and is not backed by a stud.

c. Dome The dome is a reinforced concrete shell 3'

-6" thick and 69'

-11" in radius. Due to the change in concrete thickness, the discontinuity of concrete at the springline is on the outer surface.

Reinforcing steel in the dome consists of hoop, meridional, and diagonal bars, as in the cylinder. The meridional and diagonal bars are continuous with those in the cylinder. One-half of the meridional bars, in an alternating fashion, are terminated at 60 above the springline with the remaining bars evenly spread and continued across the upper 30 of the dome.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 5 The hoop bars in the dome are terminated where no longer needed at 75 above the springline.

One-half of the diagonal bars are terminated where no longer needed, at approximately 30 above the springline; the remaining bars are terminated at approximately 45 above the springline.

See Figure 3.8-7 for details of the reinforcing steel.

The dome liner is 1/2" thick and flush with the outside face of the cylindrical liner. The anchorage system consists of tees on a 5'

-0" grid pattern. A bent stud is located in the center of each of the resulting 5'

-0" x 5'-0" panels to provide some additional anchorage. See Figure 3.8-5. d. Steel Components Steel components that resist pressure and are not backed by structural concrete include the following:

1. Equipment hatch
2. Personnel air lock
3. High energy piping penetrations
4. Moderate energy piping penetrations
5. Electrical penetrations 6. Fuel transfer tube assembly
7. Instrumentation penetrations
8. Ventilation penetrations These are discussed in Subsection 3.8.2. 3.8.1.2 Applicable Codes, Standards and Specifications The design, materials, fabrication, construction, testing and inspection of the containment structure conform to the applicable sections of the following codes and specifications which are used to establish design bases and methods, analytical techniques, material properties and quality control provisions.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 6 Dates and revision given for the listed codes are of the earliest version that was used. Subsequent issues were incorporated into the design where practicable or where the new issue directly affected the safety of the structure.

Any exceptions to the indicated issue of the ASME B&PV Code,Section III, Division 2, are indicated in the text, either in Subsection 3.8.1.4 or in Subsection 3.8.1.5, as appropriate.

Code or Specification Title ACI 211.1-74 Recommended Practice for Selecting Proportions For Normal Weight Concrete ACI 214-65 Recommended Practice for Evaluation of Compression Test Results of Field Concrete ACI 301-72 Structural Concrete for Building ACI 304-73 Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete ACI Committee Report

74-33 Recommended Practice for Hot Weather Concreting ACI Report 306R

-78 Recommended Practice for Cold Weather Concreting ACI 308-71 Recommended Practice for Curing Concrete ACI 309-72 Recommended Practice for Consolidating Concrete ACI 311-64 Recommended Practice for Concrete Inspections ACI 315-65 Manual of Standard Practice for Detailing Reinforced Concrete Structures ACI 318-71 Building Code Requirements for Reinforced Concrete (with Commentary)

ACI 347-68 Recommended Practice for Concrete Formwork ACI 614-59 Recommended Practice for Measuring, Mixing and Placing Concrete ACI Committee Report

68-33 Placing Concrete by Pumping Methods

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 7 Code or Specification Title ACI Committee Report 72-33 Placing Concrete with Belt Conveyors ACI SP2 Manual of Concrete Inspection, 1975 Edition CRSI Reinforced Concrete

- Manual of Standard Practice, 22nd Edition, first printing, 1976 CRSI Recommended Practice for Placing Reinforcing Bars, 1968 Edition ASME ASME Boiler and Pressure Vessel Code,Section III, Divisi on 2, Code for Concrete Reactor Vessels and Containments:

Concrete Containment Component

- 1975 Edition Containment Liner

- 1975 Edition (1976 Winter Addendum used for liner allowable stresses)

Reinforced Concrete

- 1975 Edition through Winter 1976 Addenda (also 1977 Winter Addendum, Article CC

-3422 and 1979 Summer Addendum, Article CC-3422) SIT - 1980 Edition, as referenced in UFSAR Subsection 3.8.1.7a, except as noted in applicable subsections of UFSAR Section 3.8.1 2 ASME ASME Boiler and Pressure Vessel Code,Section II, Material Specification, Part C, Welding Rods, Electrodes and Filler Metals (up to and including Winter 1975 Addenda)

ASME ASME Boiler and Pressure Vessel Code,Section IX, Welding and Brazing Qualifications (up to and including Winter 1974 Addenda) 1 From here on referred to as Division 2. 2 Article CA

-8000 of Division 2 applies, except that in lieu of code symbol application a New Hampshire State special waiver, dated March 18, 1976, regarding marking, stamping and recording has been granted.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 8 Code or Specification Title ASME ASME Boiler and Pressure Vessel Code,Section V, Nondestructive Examination (up to and including 1974 Summer Addendum) ASME ASME Boiler and Pressure Vessel Code,Section III, Division 2 Code Cases:

Code Case N-218, Testing Lots of Carbon Steel Solid, Base Welding Electrode on Wire, August 14, 1981.

Code Case N-219, Rules for Design of Peripheral Shear Reinforcing, January 8, 1979.

Code Case N-287, Rules for Design of Radial Shear Reinforcing, July 14, 1980.

Code Case N-232, Alternate Rules for Development Length of Reinforcing Steel Not Required to Carry Load, January 21, 1982.

ASME ASME Boiler and Pressure Vessel Code,Section II, Material Specification, Part A, Ferrous Materials (up to and including Winter 1974 Addenda)

SA-36 Specification for Structural Steel SA-240 Specification for Stainless and Heat Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure Vessels SA-300 Specification for Notch Toughness Requirements for Normalized Steel Plates for Pressure Vessels SA-333 Specification for Seamless and Welded Steel Pipe for Low Temperature Service SA-442 Specification for Carbon Steel Plates with Improved Transition Properties for Pressure Vessels SA-516 Specifications for Carbon Steel Plates for Pressure Vessels for Moderate and Lower Temperature Service

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 9 Code or Specification Title ASTM A36-70 Specification for General Requirements for Delivery of Rolled Steel Plates, Shapes, Sheet Piling and Bars for Structural Use ASTM A184-65 Specification for Fabricated Steel Bar or Rod Mats for Concrete Reinforcement ASTM A185-70 Specification for Welded Steel Wire Fabric for Concrete Reinforcement ASTM A325-71 Specification for High Strength Bolts for Structural Steel Joints, Including Suitable Nuts and Plain Hardened Washers ASTM A370-75a Methods and Definitions for Mechanical Testing of Steel Products ASTM A449-68 Specification for Quenched and Tempered Steel Bolts for Structural Steel Joints ASTM A490-71 Specification for Quenched and Tempered Alloy Steel Bolts for Structural Steel Joints ASTM A501-71 Specification for Hot Formed Welded and Seamless Carbon Steel Structural Tubing ASTM A514-70 Specification for High

-Yield Strength, Quenched and Tempered Alloy Steel Plate, Suitable for Welding ASTM A519-75 Standard Specification for Seamless and Alloy Mechanical Tubing ASTM A615-75 Specification for Deformed Billet

-Steel Bars for Concrete Reinforcement ASTM C29-71 Standard Methods of Test for Unit Weight of Aggregate ASTM C31-69 Standard Method of Making and Curing Concrete Compressive and Flexural Strength Test Specimen in the Field ASTM C33-71a Specification for Concrete Aggregates ASTM C39-71 Standard Method of Test for Compressive Strength of Molded Concrete Cylinders

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 10 Code or Specification Title ASTM C40-73 Standard Method of Test for Organic Impurities in Sands of Concrete ASTM C42-68 Standard Method of Obtaining and Testing Drilled Cores and Sawed Beams of Concrete ASTM C70-73 Standard Method of Test for Surface Moisture in Fine Aggregate ASTM C87-69 Standard Method of Test for Effect of Organic Impurities in Fine Aggregate on Strength of Mortar ASTM C88-73 Standard Method of Test for Soundness of Aggregates by Use of Sodium Sulfate or Magnesium Sulfate ASTM C109-73 Standard Method of Test for Compressive Strength of Hydraulic Cement Mortars (Using 2

-inch Cube Specimens)

ASTM C114-69 Standard Methods of Chemical Analysis of Hydraulic Cement ASTM C117-69 Standard Method of Test for Materials Finer than No. 200 (75

-uM) Sieve in Mineral Aggregates by Washing ASTM C123-69 Standard Method of Test for Light Weight Pieces in Aggregate ASTM C125-74 Standard Definitions of Terms Relating to Concrete and Concrete Aggregates ASTM C127-73 Standard Method of Test for Specific Gravity and Absorption of Coarse Aggregate ASTM C128-73 Standard Method of Test for Specific Gravity and Absorption of Fine Aggregate ASTM C131-69 Standard Method of Test for Resistance to Abrasion of Small Size Coarse Aggregate by the Use of the Los Angeles Machine ASTM C136-71 Standard Method of Test for Sieve of Screen Analysis of Fine and Coarse Aggregates ASTM C138-75 Standard Method of Test for Unit Weight, Yield, and Air Content (Gravimetric) of Concrete

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 11 Code or Specification Title ASTM C142-71 Standard Method of Test for Clay Lumps and Friable Particles in Aggregates ASTM C143-71 Standard Method of Test for Slump of Portland Cement Concrete ASTM C150-71 Specification for Portland Cement ASTM C151-74a Standard Method of Test for Autoclave Expansion of Portland Cement ASTM C172-71 Standard Method of Sampling Fresh Concrete ASTM C173-75 Standard Method of Test for Air Content of Freshly Mixed Concrete by the Volumetric Method ASTM C192-69 Standard Method of Making and Curing Concrete Test Specimens in the Laboratory ASTM C 231-71T Tentative Method of Test for Air Content of Freshly Mixed Concrete by the Pressure Method ASTM C233-73 Standard Method of Testing Air

-Entraining Admixtures for Concrete ASTM C235-68 Standard Method of Test for Scratch Hardness of Coarse Aggregate Particles ASTM C260-69 Specification for Air

-Entraining Admixtures for Concrete ASTM C295-65 Recommended Practice for Petrographic Examination of Aggregates for Concrete ASTM C309-74 Standard Specification for Liquid Membrane

-Forming Compounds for Curing Concrete ASTM C494-71 Specification for Chemical Admixtures for Concrete ASTM C496-71 Standard Method of Test for Splitting Tensile Strength of Molded Concrete Cylinders

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 12 Code or Specification Title ASTM C566-67 Standard Method of Test for Total Moisture Content of Aggregate by Drying ASTM C666-75 Standard Method of Test for Resistance of Concrete to Rapid Freezing and Thawing ASTM D75-71 Standard Method of Sampling Aggregates ASTM D297-74 Methods of Chemical Analysis of Rubber Products ASTM D412-68 Method of Tension Testing of Vulcanized Rubber ASTM D573-67 Test for Accelerated Aging of Vulcanized Rubber by the Oven Method ASTM D624-73 Tests for Tear Resistance of Vulcanized Rubber ASTM D746-73 Test for Brittleness Temperature of Plastics and Elastomers by Impact ASTM D816-55 Standard Method of Testing Rubber Cements ASTM D1149-64 Test for Accelerated Ozone Cracking of Vulcanized Rubber ASTM D2240-75 Test for Indentation Hardness of Rubber and Plastics by Means of a Durometer ASTM E96-66 Standard Methods of Test for Water Vapor Transmission of Materials in Sheet Form ASTM E329-72 Recommended Practice for Inspection and Testing Agencies for Concrete as Used in Construction (Articles 7, 8 and 9 of ASTM E329 shall not apply).

AISC Code of Standard Practice for Steel Buildings and Bridges (July 1, 1970) AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings 1969 Edition (including supplements 1, 2 and 3)

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 13 Code or Specification Title AISC Specification for Design of Light Gage Cold

-Formed Steel Structural Members, 1977 Research Council for Riveted and Bolted Structural Joints of the Engineering Foundation Specification for Structural Joints Using ASTM A325 or A490 Bolts (September 1, 1966)

ANS 7.60-1972 Standard for Leakage Rate Testing of Containment Structures for Nuclear Reactors ANSI N177 Plant Design Against Missiles (Draft Standard, April, 1974)

ANSI A58.1

-1972 Building Code Requirements for Minimum Design Loads in Buildings and Other Structures ANSI N45.2

-1974 Quality Assurance Program Requirements for Nuclear Power Plants ANSI N45.3

-1973 Electrical Penetration Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations ANSI N45.4

-1972 Leakage Rate Testing of Containment Structures for Nuclear Reacto rs ANSI N101.2

-1972 Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities ANSI N101.4

-1972 Quality Assurance for Protective Coating Applied to Nuclear Facilities ANSI N512-1974 Protective Coatings (Paints) for the Nuclear Industry AWS D1.0-69 Standard Code for Arc and Gas Welding in Building Construction AWS B3.0-41 Standard Qualification for Procedures AWS D12.1-75 Recommended Practice for Welding Reinforcing Steel, Metal Inserts and Connections in Reinforced Concrete Construction

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 14 Code or Specification Title Uniform Building Code International Conference of Building Officials, Uniform Building Code, 1973 Edition U.S. Dept. of Labor OSHA Occupational Safety and Health Administration Standards, October, 1975 Edition U.S. Dept. of Labor

QQ-C-40 Caulking: Lead Wool and Lead Pig, 1963 NRC TID 7024 Nuclear Reactors and Earthquakes (Prepared by Lockheed Aircraft Corp. for NRC, August 1963)

PS-1-74 U.S. Dept. of Commerce, National Bureau of Standards, Construction and Industrial Plywood Steel Structures Painting Manual Vol. 1 Steel Structures Painting Council, Good Painting Practice, 1966 Steel Structures Painting Manual Vol. 2 SSPC, Systems and Specifications, 1973 Federal Specifications and Standards Handbook 44 National Bureau of Standards Specifications, Tolerances and Other Technical Requirements for Commercial Weighing and Measuring Devices, 1971 U.S. Army Corps of Engineer Specifications Title CRD-C38-73 Method of Temperature Rise in Concrete CRD-C39-55 Coefficient of Thermal Expansion CRD-C44-63 Coefficient of Thermal Conductivity CRD-C119-53 Method of Test for Flat and Elongated Particles in Coarse Aggregate CRD-C-588-78A Specification for Expansive Grouts CRD-C-589-70 Methods of Sampling and Testing Expansion Grou ts SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 15 NRC Regulatory Guides Title 1.10, Rev. 1, 1/73 Mechanical (Cadweld) Splices in Reinforcing Bars of Concrete Containments 1.15, Rev. 1, 12/72 Testing of Reinforcing Bars for Concrete Structures 1.18, Rev. 1, 12/72 Structural Acceptance Test for Concrete Primary Reactor Containments 1.19, Rev. 0, 8/72 Nondestructive Examination of Primary Containment Liners 1.54, Rev. 0, 6/73 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants 1.55, Rev. 0, 6/7 3 Concrete Placement in Category I Structures 1.57, Rev. 0, 6/73 Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components 1.60, Rev. 1, 12/73 Design Response Spectra for Seismic Design of Nuclear Power Plants 1.61, Rev. 0, 10/73 Damping Values for Seismic Design of Nuclear Power Plants 1.76, Rev. 0, 4/74 Design Basis Tornado for Nuclear Power Plants 1.84, Rev. 15. 5/79 Code Case Acceptability ASME Section III Design and Fabrication 1.85, Rev. 15, 5/79 Code Case Acceptability ASME Section III Materials 1.92, Rev. 1, 2/76 Combining Modal Responses and Spatial Components in Seismic Response Analysis 1.94, Rev. 1, 4/76 Quality Assurance Requirements for Installation and Testing of Structural Concrete and Structural Steel During Construction Phase of Nuclear Power Plants 1.136, Rev. 2, 6/81 Materials, Construction, and Testing of Concrete Containments

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 16 Code of Federal Regulations Title 10 CFR 20 Standards for Protection Against Radiation 10 CFR 50 App. A General Design Criteria for Nuclear Power Plants 10 CFR 50 App. B Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR 50 App. J Primary Reactor Containment Leakage Testing for Water

-Cooled Power Reactors 10 CFR 100 App. A Reactor Site Criteria The below listed UE&C design and construction specifications applicable to the containment structure were prepared in accordance with applicable codes, quality control requirements and NRC Regulatory Guides:

UE&C Specifications Title 9763.006-1-1 General Concrete Construction, Steel Erection, and Circulating Water Pipe Installation 9763.006-5-1 Civil Testing Facility and Services 9763.006-5-4 Containment Initial Integrated Leakage Rate Test (ILRT) 9763.006-5-5 Containment Structural Integrity Test (SIT) 9763.006-10-1 Dewatering 9763.006-11-1 Foundation Waterproofing 9763.006-13-2 Containment Concrete Work 9763.006-14-1 Furnishing, Detailing, Fabricating, and Delivering Reinforcing Bars 9763.006-14-2 Installation of Reinforcing Bars in Containment Structure 9763.006-15-1 Containment Liner 9763.006-15-2 Containment Equipment Hatch and Personnel Locks

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 17 UE&C Specifications Title 9763.006-18-2 Installation of Miscellaneous Embedded Steel and Weldments 9763-006-18-4 Furnishing and Installing Embedded Steel and Miscellaneous Steel 9763.006-41-4 Furnishing of Protective Coating (Paint) System Materials and Related Services 9763.006-41-7 Field Painting of Containment Structure Interior 9763.006-69-1 Concrete Batch Plant 9763.00 6-69-3 Concrete Mixes 9763.006-69-7 Standard Concrete Mixes 9763.006-80-1 Containment Design 9763.006-80-2 Construction of Containments 9763-MPS-1 Material and Processing Requirements for Nuclear Power Plant Components 9763-MPS-3 Material and Processing Requirements for Bending of Welded Studs, Reinforcing Bars and Anchor Bolts 9763-QAS-1 Quality Assurance Administrative and System Requirements (Nuclear) 9763-RM-1 Instructions for Site Records Management System 9763-SD-15-2 Seismic Requirements 9763-WS-1 Requirements for Welding and Nondestructive Examination for Nuclear Pressure Components and Nuclear Power Piping 9763-WS-1-NE Requirements for Welding and Nondestructive Examination for Nuclear Pressure Class MC Components 9763-WS-4A Requirements for Welding and Nondestructive Examination for Nuclear Containment Structure Liner

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 18 UE&C Specifications Title 9763-WS-4B Requirements for Stud Welding and Nondestructive Examination for Nuclear Containment Structure Liner 9763-WS-4C Requirements for Mechanical Splicing and Nondestructive Examination of the Reinforcing Bars Spliced by the Cadweld Method 3.8.1.3 Loads and Loading Combinations The containment is designed to withstand all credible conditions of loading, including preoperational test loads, normal loads, severe environmental loads, extreme environmental loads, and abnormal loads. These loads are determined in accordance with Article CC

-3000 of Division 2 and are considered in the applicable service and factored load combinations to assure that the response of the structure will remain within the design limits prescribed in Subsection 3.8.1.5. Site

-related loads are also considered.

a. Design Loads The definitions of the loads used in the design of the containment include the following:
1. Preoperational Test Loads Test loads are those which are applied during the initial and any subsequent structural integrity or leak rate testing of the containment. The definitions for dead loads and live loads are those given in the Normal Load section. The following loads are also considered:

(a) Test Pressure (P t) The containment is pressurized to 115 percent of the design pressure to test its structural integrity; i.e., test pressure is 60 psig with design pressure being 52 psig.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 19 (b) Test Temperature (T t) The thermal loads occurring during the structural integrity test conditions of the containment are induced by the temperature gradient between the inside containment air temperature and the outside air temperature. The maximum and minimum temperatures considered inside the containment during the test are 100 F and 50 F, respectively. During this test, temperature outside the containment (within the Enclosure Building) will be monitored and controlled to maintain a minimum temperature of 30F; also, the average differential temperature between the inside and outside of the containment does not exceed 65 F, except the equipment hatch and personnel air locks.

2. Normal Startup, Operational, and Shutdown Loads Normal loads are those loads encountered during normal plant startup, operation, and shutdown. They include the following:

(a) Dead Loads (D) Dead loads are all permanent gravity loads including, but not limited to, the weight of the base mat, cylindrical wall, dome, internal concrete structures including fill mat, steel liner plate and structural framing, equipment, piping, cable and cable trays and miscellaneous building loads within the containment. Gravity loads from the internal structures are transmitted to the base mat through the fill mat. Included as a dead load is the buoyant force of the groundwater at El.

+20'-0" (i.e., 50'

-0" above the top of the base mat) as described in Subsection 2.4.13.5. Buoyancy from the hydrostatic loading is considered in the base mat design and containment stability analysis. (b) Live Loads (L)

Live loads are those loads which vary in intensity and/or occurrence. During normal operation there are no significant live loads on the external containment surface since the Containment Enclosure Building resists snow loads and lateral loads from soil pressure and normal wind. Live loads from the internal structures are transmitted to the base mat through the fill mat and have no significant effect on the containment shell or base mat design.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 20 (c) Operational Thermal Loads (T o) The thermal loads experienced during normal operating or shutdown conditions are generated by temperature gradients through the containment shell and by liner expansion. The gradient through the shell during these normal operating conditions varies between 120F on the inside and

-10F on the outside. In the base mat the gradient varies between approximately 97F on the inside to a constant 40F at the interface with the bedrock; the top of the fill mat is at 120F. These gradients are shown on Figu re 3.8-8. Other temperature gradients, such as 50F on the inside varying to 90F on the outside may occur, but the loads produced by them do not control the design of the concrete containment. Note also that the effect of the Containment Enclosure Building in reducing the lower bound outside temperatures was not considered in the containment design; that is, -10F was used in the evaluation of thermal loads.

(d) Operational Pipe Reactions (R o) Piping reactions transmitted to the containment during norma l operation or shutdown conditions are based on the most critical transient or steady

-state condition. The magnitudes of these loads are determined by the piping design and are included in the Design Report. (The Seabrook Station Containment Design Report is prepared in accordance with Subarticle CA

-3240 of Division 2 and is retained by the Owner in accordance with the requirements of Subarticles CA

-3100 and CA

-4832.) (e) Pressure Variation (P V) Differential pressure loads result from pressure variation either inside the containment or in the annulus between the containment and the Containment Enclosure Building. This pressure variation is produced either by atmospheric fluctuations or by HVAC equipment. The containment structure was designed to withstand a maximum external pressure of 3.5 psi (differential).

The stability of the shell was investigated under the effects of this external pressure load.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 21 3. Severe Environmental Loads Severe environmental loads are those loads that would result from external conditions which could infrequently be encountered during the plant life.

The following loads are included in this category:

(a) Wind Load (W) There is no wind load considered in the containment design due to the presence of the containment enclosure.

(b) Operating Basis Earthquake Loads (E o) These are the loads generated by the Operating Basis Earthquake, which is the earthquake that could reasonably be expected to occur at the plant site during the operating life of the plant. Only the actual dead load and weights of fixed equipment are considered in evaluating the seismic response forces. The horizontal and vertical design response spectra for the OBE are derived by applying a factor of 0.5 to the response spectra given for Safe Shutdown Earthquake (SSE), which is described below. The effects of two (2) orthogonal horizontal components and one (1) vertical component of earthquake are considered and combined by the square-root-of-the-sum-of-the-squares method. Due to the presence of the Containment Enclosure Building, there are no dynamic effects of soil on the containment.

4. Extreme Environmental Loads Extreme environmental loads are those loads which result from postulated events which are credible, but highly improbable. The following loads are included in this category:

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 22 (a) Safe Shutdown Earthquake Loads (E ss) These are the loads generated by the Safe Shutdown Earthquake, which is the maximum potential earthquake that could occur in the vicinity of the site, based on geological and historica l investigations. Dead and fixed equipment loads are described under the Operating Basis Earthquake, above. The horizontal and vertical forces on the containment are developed from the response spectra given in Figure 2.5-38 and Figure 2.5-39, the development of which is described in Subsection 2.5.2.6. The effects of two (2) orthogonal horizontal earthquakes and one (1) vertical earthquake are considered and combined by the square-root-of-the-sum-of-the-squares method. Due to the presence of the containment enclosure, there are no dynamic effects of soil on the containment.

(b) Tornado Loads (W t) Due to the presence of the Containment Enclosure Building, wind pressure and pressure variation are not considered in the containment design. In addition, the containment structure is 3'

-6" thick at the minimum, and tornado

-generated missiles are not deemed capable of penetrating a 24" thick reinforced concrete wall (see Subsection 3.5.1.4). Therefore, missile effects due to the tornado are also not considered in the containment design.

5. Abnormal Loads Abnormal loads are those generated by postulated high energy ruptures, particularly a rupture in the Reactor Coolant System resulting in a loss-of-coolant accident (LOCA). Post LOCA containment flooding is also considered. The maximum level of flooding, however, is 5'

-4" above the top of the fill mat (-26 feet); this depth of water causes negligible loading on the containment structure.

(a) Accident Pressure (P a) A transient pressure load is used for the design of the containment. The maximum calculated internal pressure associated with the DBA is 49.6 psig. This provides a margin of 4.8 percent for the design pressure which is 52.0 psig. The pressure

-transient curve for the containment is shown on Figure 3.8-9.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 23 (b) Accident Temperature (T a) The transient temperature increase of the liner was considered in the design of the containment. The maximum liner temperature is 268F. However, a maximum liner temperature of 271F has been used in the design. The temperature transient curves for the containment liner are shown on Figure 3.8-10 and Figure 3.8-11. The time-dependent thermal gradient through the concrete of the dome, cylinder and base mat was also considered in the design of the containment. When the accident pressure, P a, is considered, the coincident thermal gradient is equivalent to the normal operating gradient. Due to the high insulating properties of the concrete, the pressure peak occurs before the temperatures within the concrete are appreciably altered. For design of the cylinder and dome, the peak liner temperature and peak pressure were also considered to occur simultaneously. This produced the most conservative design condition where responses to the loads are additive. Where responses are not additive, peak pressure was considered without the thermal loads. The thermal gradients used for the design of the containment are shown on Figure 3.8-8. For the design of the liner, the transient conditions of liner temperature and coincident accident pressure were considered in order to produce the most conservative design.

(c) Accident Pipe Reactions (R a) Pipe reaction loads due to thermal conditions generated by the DBA, including R o, were considered in the design. The magnitude of these loads was determined by the piping design and is included in the Design Report.

(d) Pipe Break Loads (R r) These are local effects on the containment due to the DBA, as follows: (1) R rr = load on the containment generated by the reaction of a ruptured high energy pipe. The time dependent nature of the load and the ability of the containment to deform beyond yield are considered in establishing the structural capacity necessary to resist the effects of R rr.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 24 (2) R rj = load on the containment generated by jet impingement from a ruptured high energy pipe. In general, direct impingement of steam on the containment does not produce significant design loadings due to the distance between the wall and the break location. Where a break is postulated to occur close enough to produce a critical loading, a shield or deflector is provided, and the loading is transferred to the embedment for the pipe whip restraint to which the shield is attached.

(3) R rm = load on the containment resulting from the impact of a ruptured high energy pipe. Generally, all high energy lines are constrained by pipe restraints, and loading of this nature is prevented. However, in isolated cases where there are postulated pipe whip impact loads, liner adequacy was evaluated for such loads.

6. Site-Related Loads Site-related loads are loads peculiar to the Seabrook site. These loads are not combined with abnormal loads but are considered on an individual basis. (a) Aircraft Impact The impact of an FB

-111 type aircraft weighing 81,800 pounds and traveling at 200 mph was evaluated. The effects of this aircraft crash on the containment have been investigated and were found not to be a controlling design consideration. For a complete discussion of the aircraft impact analysis, see Appendix 2P, entitled "Seabrook Station Containment Aircraft Impact Analysis." (b) Turbine Missiles The probability that a turbine missile will strike the containment has been shown to be less than 1.0x10

-7 per year. Consequently, this type of missile was not considered in the design. See Subsection 3.5.1.3 for a complete discussion.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 25 c) Explosions and Delayed Ignition of Vapor Cloud Structural loadings from explosions and delayed ignition of a vapor cloud were shown to be of a very low magnitude and were not considered in the design. See Subsection 2.2.3.1 for a complete discussion.

(d) Flood Loads (F) The effect of the design basis flood to Elevation +20.6' (50.6' above the top of the base mat) is considered in the design of the containment. Due to the presence of the Containment Enclosure Building, the only effect of the flood is its buoyant effect on the containment of the structure.

7. Other Pipe Break Loads All high energy pipe lines are postulated to break, and all moderate energy lines are postulated to crack. The loads produced by these accidents, however, have no effect on the containment design, either because of shielding around the pipes or because of the distance separating the pipes and the containment wall.
8. Cyclic Loading The various cycles loads were considered in the design. The following design conditions were considered in the fatigue analysis:

120 cycles start and shutdown 500 OBE cycles 100 SSE cycles 1 accident cycle (LOCA) 160 pressure test cycles (equipment hatch and personnel air locks)

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 26 b. Load Combinations Various combinations of loads were used to determine the maximum strength required of the containment at various locations. These combinations were divided into service and factored load combinations.

1. Service Load Combinatio ns Service load combinations include conditions encountered during testing, normal operation, shutdown and severe environmental conditions; these are listed in Table 3.8-1. Under these conditions, the structural components are designed to remain within elastic limits and satisfy the stress limitations specified in Subsection 3.8.1.5. 2. Factored Load Combinations Factored load combinations include those conditions resulting from severe environmental, extreme environmental, abnormal, abnormal/severe environmental and abnormal/extreme environmental loads, as defined in Division 2 and listed in Table 3.8-1. For each of these loading categories the structure is designed so that the allowable stresses comply with Article CC

-3000 of Division 2, and the overall structural behavior is predicted to remain in the elastic range when thermal effects are not included. Design assumptions are presented in Subsection 3.8.1.4. 3. Additional Notes on Load Combinations The load combinations, with their appropriate load factors, which require investigation to assure that the maximum effects of all load combinations are considered, are in accordance with the requirements of Division 2 and are given in Table 3.8-1. These combinations are used in the overall design of the concrete containment and also in the design of localized areas, such as penetrations and shell discontinuities.

For loads that vary, those values which produce the most critical combination of loading were considered. The live load (L) is considered to vary from zero to full value for all load combinations.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 27 Long-term conditions, such as operating thermal loads, creep and shrinkage, which produce compression in the reinforcing steel, do not have a significant effect on the structural integrity of the containment structure, since the accident loads, which are the most significant loads, are generally resisted in tension by the reinforcing steel. In addition, the accident loads are short once

-occurring loads which will have negligible creep effects.

For the design of the liner, the load combinations in Table 3.8-1 are applicable with the exception that coefficients for all load cases are taken equal to 1.0.

Steel components of the containment shell that are pressure

-resisting but unbacked by concrete are designed in accordance with the ASME Code,Section III, Division 1, using loads described in Subsection

3.8.2. Earthquake

effects are not assumed to occur simultaneously with flooding effects since the maximum flood is not associated with an earthquake.

Effects of a thermal gradient through the concrete section are not considered where the effects of the gradient reduce the effects caused by an abnormal loading condition.

Maximum values of time

-dependent loads such as accident pressure, temperature and pipe break loads are considered.

The load combinations in Table 3.8-1 are applicable to the computations of factors of safety against overturning, sliding and flotation, with the exception that the coefficient for live load is zero. Buoyant forces are conservatively considered to decrease the dead loads for determination of overturning and sliding.

3.8.1.4 Design and Analysis Procedures The containment structure is designed as a reinforced concrete thin shell structure in accordance with the requirements of Article CC-3000 of Division 2, as described in this Subsection and in other subsections of Subsection 3.8.1, and in accordance with the other applicable codes, standards and specifications defined in Subsection 3.8.1.2. The containment structure is designed to safely withstand the load combinations as defined in Subsection 3.8.1.3, and to provide biological shielding for normal and accident conditions. The critical areas for analysis were the base mat, the intersection between cylinder wall and base mat, the liner plate system, and the penetration openings.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 28 The walls, dome and base mat of the containment are a reinforced concrete system with a leaktight steel liner attached to the inside surface, and were designed to behave as a single integrated system under the applied load combinations.

The containment structure as a whole behaves as a membrane structure, except in areas of discontinuity where there are local shear forces and moments. When subjected to internal pressure, the dome and walls displace outward with slight discontinuity effects at the dome and wall intersection. Discontinuity effects also exist near openings and at the intersection of the wall and base mat.

An iterative approach to design was taken, in which a proposed structural system was analyzed for stresses, strains and displacements using the finite element method, and then checked against design acceptance criteria, as defined in Subsection 3.8.1.5. The iteration process was repeated until an acceptable design was achieved.

The objective of the analysis of the reinforced concrete portion of the containment was the determination of maximum stresses in concrete and reinforcement over the range of boundary conditions, cracking assumptions and load combinations. To this end, the reinforced concrete was analyzed as a layered system, which allowed the investigation of each discrete layer of meridional, hoop and seismic reinforcing.

a. Design The containment was designed to retain its functional capability during normal operation and emergency conditions. To meet this criterion, the leaktight integrity of the liner is maintained, and the structure was designed to respond elastically under all mechanical loading, except as noted in Subsection 3.8.1.5. The design criteria are based on the applicable codes discussed in Subsection 3.8.1.2. Typical reinforcing details are shown on Figure 3.8-1, Figure 3.8-2, Figure 3.8-3 and Figure 3.8-4, Figure 3.8-6 and Figure 3.8-7. b. Analytical Techniques The analysis of the containment was performed by computer programs which utilize the finite element method. The finite element method was implemented by idealizing the containment with a system (model) of appropriately shaped elements which are interconnected at node points. These node points are located at the intersections of the lines which define the boundaries of each element.

Generally, two types of analyses, axisymmetric and three

-dimensional, were performed.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 29 A finite element axisymmetric analysis, using solid and shell ring elements with orthotropic and elastic material properties, was used in the determination of displacements and forces in the containment structure with temperature, pressure and dead loads. The concrete, steel reinforcement, and the steel liner were represented by a system of ring elements which are triangular, linear, or quadrilateral in shape, as shown in Figure 3.8-12. The hoop, meridional and seismic reinforcement was modeled with shell elements located near the inside and outside surfaces of the wall and dome, as shown in Figure 3.8-13. Each layer of reinforcement was idealized as an independent uncoupled orthotropic steel plate. Three-dimensional finite element analyses using elements of appropriately shaped sections were applied in the determination of displacements, stresses and strains for nonaxisymmetric loadings and local regions of the containment. Various types of elements were used for this analysis, such as shells and solids with isotropic and orthotropic material properties. The concrete, reinforcing steel and steel liner were idealized using layered elements. The three

-dimensional finite element method was applied to the analysis of localized areas of the containment. A detailed discussion of the finite element programs used in the above mentioned analyses is given in Subsection 3.8.1.4g, and Appendix 3F. c. Assumptions on Boundary Conditions

1. Containment Shell Analyses The base mat was analytically treated as an integral part of the containment to realistically represent its effect on the containment shel l

behavior. The bottom of the base mat was assumed fixed in the vertical direction but free in the radial direction in all axisymmetric analyses for pressure, temperature and dead loads. For the seismic analyses described in Subsection 3.7.2, full fixity was assumed at the base of the containment wall. Table 3.8-4 summarizes the load case and boundary condition assumptions for the containment shell analyses.

The containment mat with the reactor cavity pit and a portion of the shell was modeled in separate analyses. A discussion of boundary conditions is included in Subsection 3.8.1.41.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 30 2. Large Opening Analyses A quarter sector of the containment was modeled. This model extended from the base to the apex of the dome in the meridional directions and from a plane which passes through the opening to a plane at 90 to the centerline of the opening, as shown in Figure 3.8-14. The base was assumed fully fixed; the boundary conditions on the two sides of the model are defined as shown in Figure 3.8-14 and Table 3.8-5. These displacement boundary conditions correspond to the displacement degrees-of-freedom of the thick shell element discussed in Subsection 3.8.1.4j. 3. Aircraft Impact Analysis The aircraft impact analysis described in Appendix 2P included two elastic dynamic analyses of the containment: an axisymmetric analysis (impact at apex of dome) and an asymmetric analysis (impact at springline). A shell model of the containment was used and full fixity was assumed at the base. d. Axisymmetric and Nonaxisymmetric Loads

1. The primary behavior of the containment under axisymmetric loads is predominantly thin shell, membrane behavior except in regions near the base, springline and large openings where discontinuity moments and shears are significant. These latter effects are of localized nature and are discussed further in Subsections 3.8.1.4e and 3.8.1.4j. Thermal, pressure and dead loads were applied as axisymmetric loads to the axisymmetric containment model. These are the first four load cases listed in Table 3.8-4. The cracking patterns tabulated in the second column of the table were found by an iterative method. This is discussed further in Subsection 3.8.1.4f. The analyses were performed using the finite element computer code, Wilson I, "A Finite Element Analysis of Axisymmetric Solids Subjected to Symmetric Loads," by E. L. Wilson. This code is described in Subsection 3.8.1.4g and in Appendix 3F.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 31 Figure 3.8-13 illustrates the reinforced concrete idealization used in the model, and Figure 3.8-15 shows the shell

-type force and moment resultants which are calculated for sections along the shell meridian, and subsequently used in the design of the reinforcement. Figure 3.8-16 illustrates the axisymmetric containment model with integral mat, wall and dome used in the first four analyses of Table 3.8-4. Representative node points and section numbers are shown where the latter range from 1 at the apex of the dome to 98 at the base of the wall. At any section, the model is composed of shell and solid elements representing the steel liner and reinforcement and concrete, respectively. Material properties are assigned to each layer according to the orientation of reinforcement and cracked or uncracked state of concrete. Force and moment resultants were computed for each load case by integrating stresses through the cross section. These cross-sectional forces and moments are combined for all applicable load combinations of Subsection 3.8.1.3. 2. Nonaxisymmetric Loads Nonaxisymmetric loads applied to the containment structure, other than localized loads, such as internal missiles, are the seismic and aircraft impact loads. The seismic input and seismic analysis are discussed in Subsections 3.7.1 and 3.7.2, respectively. The aircraft impact load and analysis are described in Appendix 2P. Wind and tornado loads are not considered because of the presence of the Enclosure Building. An elastic analysis of the aircraft impact loading was made using the Wilson II finite element code, "Dynamic Stress Analysis of Axisymmetric Structures Under Arbitrary Loading," by Ghosh and Wilson, as revised in September 1975. This code is further described in Subsection 3.8.1.4g and Appendix 3F. The loading was represented by Fourier series functions applied to the axisymmetric model of the containment.

The shear forces and overturning moments due to OBE and SSE seismic loads were obtained from the seismic analysis described in Subsection

3.7.2. These

shear forces and moments were treated as effective static forces acting over the cross

-section of the containment shell. Modeling assumptions for these two load cases are shown in Table 3.8-4.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 32 3. Design of Reinforcement The force and moment resultants for the individual load cases were combined for the load combinations of Subsection 3.8.1.3 and Article

CC-3000 of the ASME Code. The provisions of Subarticle CC

-3432 of Division 2 were used in the sizing of reinforcement steel.

e. Transient and Localized Loads A "hot liner" transient load was analyzed and accounted for in the design of the reinforced concrete wall. A temperature spike was placed on the liner for two cases: (a) "hot liner" with no temperature gradient on concrete wall, and (b) "hot liner" with a temperature gradient through the wall (see Table 3.8-4). When combined with mechanical loads, the effect of thermal loadings is a redistribution of stresses and strains on the cross section. The stresses and strains in the liner are discussed in Subsection 3.8.1.4k, "Steel Liner Plate and Anchors." The aircraft impact analysis is described in detail in Appendix 2P. Three analyses were performed: two of the overall structural behavior and one localized elasto-plastic dynamic analysis. The analyses of overall behavior considered the conditions of impact on the dome (axisymmetric structure with axisymmetric loading) and impact on the springline (axisymmetric structure with unsymmetric loading). Both analyses assumed linear behavior and used the Wilson II finite element code. The asymmetric loading of the second analysis was represented by a Fourier series. Both analyses showed that yielding would occur local to the point of impact. Accordingly, a localized, nonlinear analysis was used to determine the extent of damage to the containment shell. The details of this

analysis are also found in Appendix 2P. In brief, effective mass and stiffness properties were determined for an assumed mode of collapse consisting of a circular fan yield

-line configuration.

An equivalent single

-degree-of-freedom nonlinear model was then subjected to an idealized force

-time loading function and the maximum deformations determined. It was shown that the "as designed" containment structure with Enclosure Building can withstand impact of an FB

-111 aircraft at 200

-mph impact speed without collapse or impairment of the leak

-tight integrity of the liner.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 33 f. Creep, Shrinkage and Cracking of Concrete Due to low sustained concrete stress associated with conventionally reinforced concrete structures, the effects of concrete creep are negligible. Since the load combinations which control the rebar design involve accident pressure, which effectively cracks the concrete and places the reinforcement into tension, creep and shrinkage

-induced stresses are not limiting factors in design. In addition, the structural integrity test cracks the concrete thereby relieving creep and shrinkage stresses that occur subsequent to construction. Therefore, the effects of creep and shrinkage were not considered in the analysis and design.

Since it was assumed that concrete has no tensile capacity, a cracked section, iterative approach was used for the analysis under the thermal, pressure and dead loads. The procedure for determining the crack pattern involved a check of the stresses in each concrete layer of the cross section. Where concrete stresses were in tension, the elastic properties in that direction (modulus of elasticity and associated Poisson's ratio) were set equal to zero. The procedure was repeated until there were no significant changes in moments and forces on the containment sections in two successive iterations. Since the finite element representation of the cross-section modeled both concrete and rebar using a layered model, as shown in Figure 3.8-13, an accurate portrayal of cross section stiffnesses under various configurations of cracking was possible. Table 3.8-4 defines the cracking pattern used for each load case. The analysis of the containment subjected to the nonaxisymmetric seismic loads considered the concrete to be uncracked. For this load case, the magnitudes of flexural moments are small compared to the flexural moments resulting from the axisymmetric loads. Hence, the redistribution of moments due to cracking under seismic loading does not control the design.

g. Description of the Computer Programs Utilized in the Design and Analyses The computer programs used in the containment design and analysis are briefly described in this subsection. A summary of the comparisons of results used to validate them is given in Appendix 3F.

The programs are either of two types: a recognized program in the public domain with sufficient history of use and documentation to justify its applicability and validity without further demonstration, or a program which gives solutions to a series of test problems that have been demonstrated to be substantially identical to those obtained from classical solutions and/or analytical results published in technical literature.

Utility programs used to replace hand calculations are not discussed. These programs were validated by comparison to sample hand calculations whenever used in the analysis.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 34 1. STARDYNE, Static and Dynamic Structural Analysis System, by Mechanics Research, Inc., 9841 Airport Boulevard, Los Angeles, CA.,

90045. Documentation is available from Control Data Corporation (Publication No. 76079900). The STARDYNE system is designed to analyze linear elastic structural models for a wide range of static and dynamic problems.

2. MARC-CDC, Nonlinear Finite Element Analysis Program, by Dr. Pedro Marcel and Associates of the MARC Analysis Corporation, 260 Sheridan Ave., Palo Alto, CA., 94306. Documentation is available from Control Data Corporation. MARC

-CDC provides elastic, elastic

-plastic, creep, large displacement, buckling and heat transfer analysis capabilities. It also performs dynamic analysis by the modal or direct integration procedures.

3. WILSON 1, (SAG 001) Finite Element Analysis of Axisymmetric Solids Subjected to Axisymmetric Loads, by E. L. Wilson of the University of California, Berkeley, July 1967

-Revised, November 1969. Documentation is available from the Earthquake Engineering Research Center of the University of California, Berkeley. The Wilson 1 computer program is based on the finite element direct stiffness method, and is applied to the determination of stresses and displacements in axisymmetric structures (solids and/or shells of revolution) subjected to axisymmetric mechanical loads or temperature gradients. The theoretical basis of the program is the work of E. L. Wilson, References 1 and 2. 4. WILSON 2, (SAG 010), Dynamic Stress Analysis of Axisymmetric Structures under Arbitrary Loading, by S. Ghosh and E. L. Wilson of the University of California, Berkeley, September, 1969

-Revised September 1975. Documentation is available from the Earthquake Engineering Research Center of the University of California, Berkeley, Report, No. EERC 69-10. The Wilson 2 program is based on the finite element stiffness method and is applied to complex axisymmetric structures subjected to any arbitrary static or dynamic loading or base excitation. The three-dimensional axisymmetric continuum is represented as an axisymmetric thin shell, a solid of revolution, or as a combination of both.

5. LESCAL calculates the stresses and strains in rebars and/or concrete in accordance with the criteria set forth in Subarticle CC

-3511 of Division

2. The section may be reinforced with horizontal, vertical and diagonal rebars. The applied loads are axial forces and moments in the vertical and horizontal faces and in

-plane shear. When in

-plane shear forces are included, a solution is obtained by solving Duchon's equations, Reference 3.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 35 6. SAG 054, Amplified Floor Response Envelope. This program generates an envelope for amplified response spectra, spreading the peaks by a user-specified amount.

7. SAG 058, Response Spectra. This program calculates the response spectra of a single degree

-of-freedom damped oscillator due to a transient base motion. The input base motion may be an arbitrary forcing function. The output consists of the maximum relative displacement, the maximum relative velocity and the maximum absolute acceleration for the various selected frequencies and the times when these values occur.

8. TAPAS, (SAG 008), "Transient Temperature Analysis of Plane and Axisymmetric Solids," Reference 4, was developed to determine the temperature distribution through a solid body as a function of time when subjected to temperature variation or heat flux inputs. A finite element technique coupled with a st ep-by-step time integration procedure is used. Both steady

-state and transient heat flow can be treated.

9. SAG 017, Fourier Coefficient Expansion Program, was developed to be used in conjunction with the Wilson 2 program to compute Fourier series representation of general nonaxisymmetric load functions.
10. SAG 024, MMIC, calculates weight, weight moments of inertia and plan location of the center of weight of a segment of a structure given the dimensions, density and location of each structural component and the magnitude and location of all concentrated loads.
11. SAG 025, SECTION, calculates beam section properties of structures for use in lumped mass stick models for dynamic analysis.
h. Tangential Shear The design of the containment for seismic loads incorporates diagonal shear reinforcing to resist tangential shear. The design for tangential shear is in accordance with the provisions of Article CC

-3000 of the ASME Code, with the exception of the limitation on allowable concrete shear stress shown as Note 1 of Table 3.8-2.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 36 The forces and moments due to seismic loading are discussed in Subsection

3.7.2. Three

components of the seismic motion were assumed to exist simultaneously, and the resulting component forces and moments were combined by the squar e-root-of-the-sum-of-the squares (SRSS) method. The maximum tangential shear from the SRSS combination is assumed to act simultaneously at all points on the containment circumference at the given elevation. All forces and moments were combined per the specified load combinations. The LESCAL program, described in Subsection 3.8.1.4g, was used to calculate rebar stresses for all sections and elevations combining the stresses due to in

-plane forces and moments. Duchon's equations (Reference

3) are incorporated into LESCAL for calculating rebar stresses (including inclined rebar) for the combined membrane forces. i. Variation in Physical Material Properties The effects of variations in material properties were considered in the design and analysis. Material properties which can strongly influence both analysis and design due to variability or uncertainty include: (1) dynamic modulus of soils, (2) the modulus of elasticity of concrete and, (3) material strengths.

As this containment is founded on rock, the first of these sources of variability is removed from consideration. The modulus of elasticity of concrete is a function of concrete compressive strength which in turn is typically substantially higher in the "as-built" structure than assumed for analysis and design. While variability in concrete modulus has no significant effect on structural design, it influences structural stiffness and natural frequency, and, subsequently, the amplified response spectra of the seismic analysis. This impacts equipment design as discussed in Subsection 3.7.3. The variability was accounted for by peak spreading when generating envelopes of the response spectra. Variability in material strength is taken into account in Division 2, Subarticle CC

-3400, design allowables.

j. Thickened Penetrations The equipment hatch and personnel air lock (27'

-5" and 7'-0" diameters, respectively) are large openings with thickened bosses. Figure 3.8-3 and Figure 3.8-4 are drawings of typical wall sections with their reinforcement configurations. The wall thicknesses and reinforcement were designed for the stress concentration effects induced by the presence of the openings in the cylinder. The main hoop and meridional reinforcement is bent around the opening to provide continuity. Additional local reinforcement is provided, including stirrups and tiebacks for bent bars.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 37 The analyses of the large openings were by the MARC finite element computer code using a three

-dimensional model of a quadrant of the containment cylinder and dome, Figu re 3.8-17. The two openings are sufficiently separated to allow independent analyses. Therefore, two planes of symmetry were assumed and the single quadrant model resulted. Appropriate symmetry and antisymmetry in loads and boundary conditions were defined. Figure 3.8-14 and Table 3.8-5 show the boundary conditions for the various load cases. The model extends from the base to the apex of the dome using a thick

-shell superparametric element (element 22 in the MARC program element library). This element incorporates eleven through-thickness layers to accurately represent the liner, reinforcement and concrete. Figure 3.8-18 illustrates the partitioning of a typical cross

-section into the 11 layers. The tapered

-thickness transition region between the normal 54-inch wall and the thick boss is modeled by stepping the element thicknesses through this region. The thick

-shell element gives all stresses, except the through-thickness (normal to wall) direct stress, at each of the eleven layers. The conventional shell-type force and moment resultants, including transverse shears, are calculated from these stresses and were used as the basis for design of the reinforcement.

Design of the reinforcement was accomplished in conformance with Division 2 provisions for load combinations and material stress allowables, as described in Subsections 3.8.1.3 and 3.8.1.5, respectively.

k. Steel Liner Plate and Anchors
1. Design The liner is anchored to the reinforced concrete with embedded stiffeners and anchors. Typical liner details are shown in Figure 3.8-5. To maintain the leak-tight integrity of the liner under service loads and factored loads, the liner was designed to follow the major strain pattern of the surrounding concrete. The liner plate was thickened in areas around penetrations as required. The stresses and strains in the liner, determined by the analysis, are within the design allowables given in Table CC

-3720-1 of Division

2. The anchorage system was designed so that it can accommodate the

in-plane loads or deformations exerted by the liner plate and/or loads applied normal to the liner surface. The anchorage system was designed so that a progressive failure of the anchorage system is precluded in the event of a defective or missing anchor. The design of the anchors considered the effects as indicated in Subarticle CC

-3810 of Division

2.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 38 The displacements and forces in the anchorage system obtained by analysis are not allowed to exceed the allowables given in Table

CC-3730-1 of Division

2. Penetrations are provided with an anchorage system capable of transferring thermal loads, pressure loads and other mechanical loads, such as piping reactions, to the concrete containment. The penetrations, brackets and attachments are designed in accordance with the provisions of the ASME Code,Section III, Division
2. Designing those portions of the penetrations which fall within the jurisdiction of Division 1, that is, those portions not backed by concrete, is described in Subsection 3.8.2. 2. Analysis The stresses and strains in the liner due to dead load, internal pressure, a "hot" liner and a temperature gradient were obtained from the containment analyses using an axisymmetric finite element model, as discussed in Subsection 3.8.1.4d. These analyses considered the liner to be integral with the concrete structure. The liner anchors were analyzed assuming the unbuckled liner remained elastic under all conditions. Construction and material imperfections described in Division 2, Subarticle CC

-3810, were considered in the liner anchor analysis including the possibility of a buckled liner panel. Liner stresses were assumed for the combined load cases. These stresses produced liner/anchor displacements in the direction of the buckled panel. An analytical model was developed in which the buckled panel was assumed to be elastic

-perfectly plastic and was represented by an equivalent non

-linear spring whose stiffness was determined by a separate nonlinear analysis using the MARC

-CDC finite element code (Subsection 3.8.1.4g). The anchors were also represented by equivalent nonlinear springs whose force

-displacement properties were determined by testing. The analytical model yielded forces and displacements of all anchors.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 39 l. Containment Mat The containment mat, the reactor cavity pit and 55' of the cylindrical containment shell above the mat were modeled with Element 22 in the MARC

-CDC finite element program. A portion of containment shell in addition to the reactor cavity pit was included in the model to represent the stiffening effect of the containment shell on the mat. The internal structure and the fill mat were not modeled, but the loads which they transmit to the structural mat were considered. The stiffening effect of the dome and the cylindrical portion of the shell above the top of the model was considered by providing additional constraint equations which enforce the plane section remaining plane assumption. In effect, these equations insure that the top of the model remains circular and lies in a plane after deformation.

The rock foundation supporting the containment structure was modeled using the continuous foundation option which is available in the MARC

-CDC program. This option requires as input foundation properties in both tension and compression at 9 integration points on the surface of each foundation element. To account for lift off, a zero stiffness was specified in tension. A very large stiffness is given to represent the rigid behavior of the rock foundation in compression. There were no vertical boundary conditions prescribed at node points on the mat. The consistent formulation in the element automatically accounts for this vertical restraint in compression. The stiffness of the internals and fill mat was not considered since there is no structural connection between these structures and the structural mat. Because of symmetry, only one

-half of the structure was modeled, (Figure 3.8-19). The following boundary conditions were applied to the model:

1. All node points in the global 134 xz plane have symmetry boundary conditions as given below:

u y x z = 0 where: u y = translation in the global y

-direction x = rotation about the global x

-axis z = rotation about the global z

-axis The rotational boundary conditions are input into the program in a local coordinate system.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 40 2. All node points at the outside edge of the mat within +

50 from the direction of the global x

-axes have been restrained in the global x-direction to account for the lateral restraint provided by the rock belo w grade. (The earthquake forces are assumed in global x

-direction.) Thus, all horizontal loads are resisted in bearing through the reactor cavity pit and a circular arc at the elevation of the centerline of the mat. The boundary conditions used in the mat analysis differ from those used in the axisymmetric analyses since the mat is represented by a three

-dimensional model rather than the two

-dimensional axisymmetric model. This permitted a more accurate representation of the boundary effects. The mat was modeled in the axisymmetric analyses only to simulate the mat restraining effect on the containment shell.

The mat was analyzed for dead, live, pressure and seismic loads. Seismic loads were applied as inertia forces which were calculated from absolut e accelerations computed in the seismic analysis, Subsection 3.7.2. At the top of the model, external forces were applied which represent the internal forces at that elevation for each load combination. Dead load, pressure, and vertical seismic loads were assumed uniformly distributed, and the moments and shears from the horizontal earthquake were distributed as normal and shear forces which vary as cosine and sine functions, respectively.

The nonlinear effects resulting from uplift required an iterative analysis. Five iterations were required to reach the equilibrium state.

When combining seismic loads with other loads, due to the inherent nonlinearity associated with mat uplift, the response due to the vertical component of the earthquake has been added algebraically to that of one horizontal component. Since the final stress state in a nonlinear problem can only be determined when all loads are applied simultaneously, it is not possible to separate out the individual contributions to the total response of two horizontal components and a vertical component of an earthquake in order to perform the SRSS. Consequently, only one horizontal component and the vertical component were applied and no SRSS process was used. The horizontal component was assumed to act along the longer direction of the reactor cavity pit (x

-direction).

m. An ultimate capacity analysis of the containment structure for internal pressure loads was performed. The pressure

-retaining capacity of the overall containment structure and the localized areas was determined. The analysis was based on actual material properties.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 41 n. A Design Report (9763.102

-CDR-1) of the containment structure was prepared in accordance with the requirements of Division

2. This report contains sufficient information to substantiate that the design of the containment structure is in accordance with the requirements of the containment design specification and the ASME codes.

3.8.1.5 Structural Acceptance Criteria

a. General The containment structure, including liner and penetrations, was designed to remain within elastic limits under service load conditions and under the mechanical loads of the factored load conditions. With thermal loads included, the reinforcing steel yielded in some regions but was within the allowable strain limit of 2 x y. Gross deformations of the containment were also checked to assure that there is no interaction with other structures or components.

The design limits imposed on the various parameters that serve to quantify the structural behavior and provide a margin of safety are in compliance with Article CC-3000 of Division

2. The allowable limits on these parameters, for service and factored loads, are given in Table 3.8-2. b. Concrete The allowable compressive stresses, including membrane, membrane plus bending and localized stresses, and shear stresses under service loads and factored loads are as specified in Article CC

-3400 of Division 2, with the following exceptions:

1. The allowable shear stress, v c, to be resisted by the concrete will not exceed 40 psi and 60 psi for load combinations 7 and 8, respectively, in Table 3.8-1. 2. The allowable concrete stresses or radial shear are based on Code Case No. N-287. 3. The allowable concrete stresses for peripheral shear are based on Code Case No. N

-219.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 42 c. Reinforcing Steel The stress and strain limitations for reinforcing steel, under service loads and factored loads, are as specified in Subarticles CC

-3432 and CC

-3422 of Division 2, respectively.

Stress concentrations normally occurring around openings and penetrations were controlled by the use of additional reinforcing steel, which resulted in stresses and strains within the above indicated limits.

There is, however, some yielding of the seismic (diagonal) reinforcing under the mechanical loads of the abnormal/severe environmental and abnormal/extreme environmental loading conditions at the edges of the transition regions below the equipment hatch and personnel air lock adjacent to what may be considered the membrane region. The net strain at these locations, with thermal load included, is less than 2y (.00414 in/in), which is the limit established in Subarticle CC-3422.1 (d) of Division 2 (1977 Winter Addendum). The structural integrity of the containment is not affected, however. The strain limit of 2y insures that the yielding under thermal load does not result in concrete cracking which would cause deterioration of the containment.

d. Liner Plate and Liner Anchorage System Tensile and compressive stress/strain limits in the liner plate, including membrane and membrane plus bending conditions, are in accordance with Subarticle

CC-3700 of Division

2. (For the liner plate, the 1976 Winter Addendum of Division 2 is used, and for anchorage system, the January 1, 1975 Edition of Division 2 is used. All supporting documentation for material procurement and fabrication activities that commenced prior to July 1, 1975, included Certificates of Conformance to the technical requirement of this edition.)

The allowable values for forces and displacements of liner anchors embedded in concrete are based on test data, and are in accordance with the limits of Table

CC-3730-1 of Division

2. Tests were performed to obtain the shear load

- displacement relationship of liner anchors. The details of liner anchor load test procedure and results are included in Appendix 3G.

The stresses and strains in penetration assemblies, brackets and attachments are also in accordance with allowables given in Subarticle CC

-3700 of Division

2.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 43 e. Stability Acceptance criteria for stability against overturning, sliding and flotation are given in Subsection 3.8.5.5. 3.8.1.6 Materials, Quality Control and Special Construction Techniques Materials used for the containment include concrete, reinforcing steel, liner plate steel and attachments thereto, and coatings, the requirements of which are in compliance with Article

CC-2000 of Division 2 and the applicable NRC Regulatory Guides listed in Subsection 3.8.1.2. Quality control procedures employed for the fabrication and construction of the containment are in compliance with Articles CC

-4000 and CC

-5000 of Division 2 and with the applicable NRC Regulatory Guides listed in Subsection 3.8.1.2. For all materials, suppliers were required to furnish Certified Materials Test Reports (CMTR), prepared in accordance with the requirements of Subarticle CC

-2130 of Division

2. A CMTR includes results of all required chemical analysis, physical tests, mechanical tests, examinations including radiographic film, repairs and heat treatments performed on the material.

Materials used in construction, as well as their respective quality control procedures, are further described in the sections that follow. Engineering properties are given in Table 3.8-3. Construction tolerances are in accordance with the criteria of Division

2. There are no special construction techniques.
a. Concrete The ready-mixed concrete which was used is a dense durable mixture of sound coarse aggregate, fine aggregate, cement, water and admixtures, in accordance with the material, proportioning, mixing, transporting, placing and testing requirements of Division
2. Testing which was performed on the concrete and on its components is described in UE&C Specification 9763.006 1, "Concrete Batch Plant." Details of the placing procedures, including hot and cold weather precautions, are described in UE&C Specification 9763.006 2, "Containment Concrete Work." In the paragraphs which follow, however, a brief description of the material is presented.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 44 Except as noted below, the containment is constructed of concrete which has a standard compressive strength at 28 days of at least 3000 psi. The base mat, reactor pit, bottom 10 feet of the cylinder, and the regions in the cylinder near the equipment hatch and personnel air lock, in which the reinforcing anchor plates are located, are constructed of concrete which has a standard compressive strength at 28 days of at least 4000 psi. To insure that these strengths were attained, verification testing was performed in accordance with the requirements of Division 2. The nominal density of the reinforced concrete was considered as 150 lbs/ft

3. The mix proportions were established on the basis of laboratory trial batches which were designed by the testing laboratory per ACI 211.1

-74. The concrete was produced under controlled conditions by a fully automatic central batch plant located on the site to assure that the proper ingredients and proportions determined by the design mix were achieved.

The cement conforms to the requirements of ASTM C150, "Specification for Portland Cement," Type II, having a low alkali content and a moderate heat of hydration.

The coarse and fine aggregates conform to the requirements of ASTM C33, "Specification for Concrete Aggregates" and the additional requirements of Subarticles CC

-2222 and CC

-3421.5.1 of Division

2. The maximum coarse aggregate size in the containment is 11/2", except in congested areas where a 3/4" (size #7, as specified in ASTM C33) maximum aggregate size mix was used to accommodate proper placement of the concrete.

The mixing water conforms to the requirements of Subarticle CC

-2223 of Division 2. To assure a plastic and workable mix, increase durability, and increase ease of placement in congested areas, admixtures were used in the concrete mix design. These admixtures consisted of air entraining agents, water reducing agents, and retarding agents. Their effects on the strength of the mix were considered in the mix design, such that the properties described previously were the properties which were obtained after the inclusion of admixtures.

Admixtures containing chloride ions were not used in the concrete for the containment.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 45 The maximum slump permitted in mass concrete for the containment was 3", except in congested areas where a 4" slump was allowed to accommodate proper placement, with slumps greater than 4" but not more than 6" (Special High Slump Concrete) used in highly congested areas. The maximum slump for concrete utilizing a superplasticizer (high range water reducer) was 8" (9" on a case-by-case basis). The maximum slump permitted for all other concrete was 4".

No aluminum materials were used in the mixing, handling, storing, transporting, or placing of concrete materials or mixes, nor were any aluminum embedments used. The maximum concrete mix temperature during placement was 80 F. All concrete operations during cold weather conditions followed the practice defined in ACI 301 and 306R

-78 except that concrete as placed shall not be lower than 45F. Concrete was maintained at 50 F. During cold weather curing of the concrete, concrete surfaces whose temperatures are below 50F by accident for short periods of time, but remain 40F or above, have had the 7

-day curing period extended by the amount of time the concrete was below 50F (rounded out to the nearest whole day).

Alkali Silica Reaction is discussed in subsection 3.8.4.6.

b. Reinforcing Steel Reinforcing steel consists of high

-strength deformed billet steel bars conforming to ASTM A615, Grade

60. This steel has a minimum yield strength of 60,000 psi, a minimum tensile strength of 90,000 psi and a minimum elongation of 7 percent in an 8" gage length.

In addition to the Certified Material Test Reports, user tests, as required by Division 2 and Regulatory Guide 1.15, were performed by the Material Manufacturer on full

-size diameter test specimens to further verify the physical properties of the rebar.

Arc welding of rebar was not permitted.

All reinforcing bars were detailed by the Manufacturer in accordance with the requirements of the Design Drawings. Detail drawings were reviewed by the Designer.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 46 All reinforcing bars were spliced in accordance with Division 2 and UE&C Specification 9763

-WS-4C. No. 14 and 18 bars were joined by mechanical butt splices (Cadweld splices). These splices met the requirements of Subarticle CC-4333.4.4. of Division

2. The splice sleeve material conformed to ASTM A519 (85 ksi

- Minimum Ultimate Tensile Strength). Cadweld anchor plates used as mechanical anchorage for the terminated rebars, predominately in the equipment hatch and personnel air lock areas, conformed to SA537, Class

1. The splices were tested in accordance with the requirements of UE&C Specification 9763

-WS-4C to assure that they develop the specified strength. This test program is in accordance with the requirements of Division 2 and NRC Regulatory Guide 1.15, as discussed in Section 1.8, Regulatory Guide 1.136. c. Containment Liner The steel liner plate is carbon steel conforming to ASME SA 516, Gra de 60. This steel has a minimum yield strength of 32,000 psi and a minimum tensile strength of 60,000 psi, with an elongation of 21 percent in an 8" gauge length, when tested to failure.

The materials, standards and quality control procedures used for the liner and penetrations are in accordance with Division 2 and NRC Regulatory Guide 1.19. Inspections are made in accordance with Article CC

-5000 of Division

2. The following materials were used for the penetration sleeves, equipment hatch and personnel locks: 1. Equipment Hatch

-ASME SA-516, Gr. 60 Normalized

2. Personnel Locks

-ASME SA-516, Gr. 70 Normalized

3. Fuel Transfer Sleeve

-ASME SA-240, Type 304 Stainless Steel

4. Seamless Penetration Sleeves

-ASME SA-333, Gr. 1

5. Rolled Penetration Sleeves

-ASME S A-516, Gr. 60 Normalized.

Permanent attachments to the liner were fabricated from SA

-36 or SA-516 Grade 60 or SA-516 Grade 70 plate. The anchorage system, tees, brackets and attachments to the liner plate were ASME SA

-36. All backing strips were of the same material specification as the item being welded.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 47 All welding materials conformed to the requirements of UE&C Specification No.

9763-WS-4A and ASME Section II, Material Specification, Part C

- Welding Rods, Electrodes and Filler Metals.

Studs and stud welding materials were in accordance with the requirements of Subsection CC-2620 of Division

2. All liner materials are normalized, and sections in excess of 5/8" thickness were impact tested to 15 ft

-lbs. at 20F in accordance with the requirements of Division 2. For penetrations, the equipment hatch and personnel air lock, the test temperature was

-10F in the heat affected zone and

-25F in the parent metal.

When required, welds were post

-weld heat treated in accordance with the requirements of Division 2. The portions of the equipment hatch and personnel air lock within the jurisdiction of Division 1 were designed and detailed by the Fabricator. These portions are described in Subsection 3.8.2. d. Steel and Concrete Coating System The coatings used on steel, other than stainless steel, and concrete surfaces inside Containment, that might be exposed to spray have been tested and accepted in accordance with the requirements of ANSI Standard N101.2, except in the acceptance criteria used, and were applied in accordance with the manufacturer's printed instructions over properly prepared surfaces. The acceptance criterion for power tool cleaning methods, intended for limited use in the containment, is adherence of the coating (no solid debris generated) rather than specific ANSI blister size and frequency. (See also Section 6.1 for further discussion on BOP and NSSS equipment and structures coating systems inside Containment.)

Ferrous surfaces to which coatings were applied were abrasive blast cleaned in accordance with the Steel Structure Painting Council Specifications SP6 and SP10, except for the limited use of qualified power tool cleaning methods where blast cleaning became impractical. All horizontal and vertical concrete surfaces which were coated were washed and neutralized just prior to the initial coating application in order to produce a clean contamination

-free surface. All concrete patching work, such as pointing of form tie holes and removal of sharp edges and fins, was completed before the washing was begun.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 48 e. Grout Sand Cement Grout used in general repair and patch work was in conformance with the requirements of ASME Section III, Division 2 Code, 1975 edition including Winter 1975 Addenda Subarticle CC

-2240. Prepackaged cement grout used in general repair and patch work met the requirements of ASME Section III, Division 2 Code, 1975 edition including Winter 1975 Addenda Subarticle CC

-2240 except as modified herein. Aggregates met the requirements of ASTM C33 except that the gradations were adjusted as required to meet the Material Manufacturer's requirements for the product's applications.

The ASTM C88 Standard Test method for soundness of aggregates was not required to be performed provided alternate testing was performed by the Material Manufacturer to assure that the grout provided adequate resistance to weathering action. The Material Manufacturer provided a notarized Certificate of Conformance (C of C) and Certified Material Test Reports (CMTRs) which contain the results of tests required in Corps of Engineers' CRD

-C-589-70 and CRD

-C588-78A. Compressive strength of grout was tested in accordance with CRD

-C-588-78A, Section 10.3 "Compressive Strength" except that nine (9) cubes were made from each test mixture, and cubes were tested for compressive strength at 3, 7, and 28 days. CMTRs and Master Builders 713 Grout did not include chemical analysis.

f. Epoxy Bonding Materials Sikadur High

-Modulus Epoxy Bonding Adhesive (Sikastix 370) was used in general concrete repair and patching work when directed by the Construction Manager. The bonding adhesive was stored, mixed and applied in accordance with the Material Manufacturer's instructions.

The Material Manufacturer provided a notarized Certificate of Conformance for each batch of material supplied.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 49 3.8.1.7 Testing and In

-Service Inspection Requirements The structural testing and in

-service inspection program consists of the integrated leak rate test, in-service leak rate testing, the preoperational structural integrity test, and general visual inspection of structurally critical areas.

The in-service leak rate testing of the containment and visual inspection requirements are discussed in Subsection 6.2.6. The preoperational structural integrity test and visual examination are described below.

Since no new or previously untried design approaches are used for the containment structure, there are no special testing or in

-service surveillance requirements.

a. Structural Integrity Test To demonstrate that the concrete containment structure will respond satisfactorily to the postulated internal pressure loads, a preoperational structural integrity test (SIT) will be performed at 1.15 times the containment design pressure of 52 psig and at an average differential temperature between the inside and outside of the containment (within the Enclosure Building) not to exceed 65 F. The SIT will be conducted in accordance with the nonprototype requirements of Article CC

-6000 of Division

2. The Consolidated Edison's Indian Point Unit No. 2 and Washington Public Power Supply System Unit No. 1 are the prototypes for this test. Structural acceptance is based on gross deformations (diameter change, vertical growth and radial growth at the equipment hatch), concrete crack widths, deflection recovery and post test visual examination of concrete and liner. Prior to the test, a table of predicted gross deformations, crack widths, etc., will be provided as a guide for verifying satisfactory structural response during the test.

These acceptance criteria will be as follows:

1. No yielding of reinforcement as determined by analysis of crack width and deflection data.
2. No visible signs of permanent damage to the structure or liner.
3. The measured maximum deflections at points of maximum predicted deflection shall not exceed predicted values by more than 30 percent. This requirement will be waived if the 24

-hour recovery is greater than 80 percent. 4. The deflection recovery 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after complete depressurization shall be a minimum of 70 percent.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 50 The instrumentation required to obtain the data needed to verify the structural response will include methods to measure radius and diameter deformation, vertical deformation, deformation around the large equipment hatch opening and breech

-type personnel air lock opening, and crack measurements as required by ASME Section III, Division

2. The data will be obtained without embedding any devices in the cylinder or dome. A test report will be provided which will compare test results with predicted and allowable limits and evaluate any deviations.

Before and after the test, the containment will be visually examined to assure that no distress has developed on the concrete or liner.

3.8.2 Steel

Containment The following sections contain the physical descriptions, applicable codes, standards and specifications, loads and load combinations, design and analysis procedures, allowable stresses, quality control and testing requirements for the major steel penetrations of the concrete containment structure that are pressure

-resisting but unbacked by concrete. The containment structure itself is constructed of reinforced concrete and, as such, is discussed in Subsection 3.8.1. 3.8.2.1 Description of the Containment Penetrations The containment penetrations described in this Subsection are the personnel air lock and equipment hatch, the fuel transfer tube assembly, the piping, electrical and instrumentation penetrations, and the ventilation penetrations. These components penetrate the containment shell to provide access, anchor piping, or furnish some other operational requirement. They also maintain leak tightness for the containment shell, and meet the maximum allowable leakage rate, as described in Subsection

3.8.1. Other

functional characteristics are described below in the descriptions of individual components.

All penetrations are anchored to sleeves (or to barrels) which are embedded in the concrete containment wall. This embedment is accomplished by means of an engineered anchorage system that is welded to the sleeve (or barrel) which is, in turn, welded to the locally thickened liner (see Subsection 3.8.1). a. Personal Air Lock (Breech Type)

The personnel air lock (Figure 3.8-20) consists of the air lock doors and the lock barrel. Its centerline is located at elevatio n 29'-6" and an azimuth of 315, as shown on Figure 1.2-4.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 51 Significant dimensions are as follows:

Parameter Dimension Clear Opening 7'-0" O.D. of Flange on Door 7'-9 1/8" Barrel Thickness 5/8" Cover Thickness (Spherical Dished Head) 5/8" The air lock barrel has a door on each end, each of which is designed to withstand the design pressure from inside the containment. The doors are hinged and swing away from the air lock barrel. Each door is fitted with two seals that are located so that the area between seals can be pressurized and tested to 52.0 psig. A leak chase system is provided over the barrel

-liner joint of the personnel air lock for leak testing.

The locking device for the doors is a rotating, third ring, breach

-type mechanism. These doors are interlocked so that only one door can be opened at a time. The capability exists for bypassing this interlock and relieving the internal pressure by use of special tools. The doors may be operated mechanically.

A sight glass is provided at the exterior of each door to permit observation of the status of the opposite door.

Separate limit switches are provided to allow remote indication of door position, and a signal is furnished for use by the control room.

The barrel, which is also the sleeve for the personnel air lock, is embedded and anchored in the shell of the concrete containment.

b. Equipment Hatch The equipment hatch (Figure 3.8-21) consists of the barrel, the spherical dished cover plate with flange, and the air lock mounting sleeve. The centerline of the hatch is located at elevation 37'

- 0 1/2" and an azimuth of 150, as shown on Figure 1.2-4. The hatch opening has an inside diameter of 27'

- 5". A sleeve for a personnel air lock, the inside diameter of which is 9'

-10", is provided at centerline elevation 30'

-6". Thicknesses of the primary components are as follows:

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 52 Component Thickness Barrel 3 1/2" Spherical Dished Cover Plate 1 3/8"" Flange 5 3/8" Air Lock Mounting Sleeve 1 1/2" The equipment hatch cover is fitted with two O

-ring seals that enclose a space which can be pressurized and tested to 52.0 psig. The flange of the cover plate is attached to the hatch barrel with 32 swing bolts, 1 3/8" in diameter. A leak chase system is provided over the barrel

-liner joint of the equipment hatch for leak testing. The barrel, which is also the sleeve for the equipment hatch, is embedded in the shell of the concrete containment.

Provision has been made to lift the equipment hatch cover to the side to clear the opening and to store the cover in the storage saddles designed for this purpose.

A platform to allow access to the swing bolts at the top of the hatch has been permanently installed on the equipment hatch. Extensions have been added to the lower lifting lugs. These extension legs make the lifting device attachment points at approximately the same elevation for all four lug locations. The extension legs and the upper lifting lugs are used to support the hatch access platform. This platform is provided with fall protection chain railings and safety harness tie

-off points. Inserted into the mounting sleeve through the equipment hatch cover is a personnel air lock consisting of two air lock doors, two air lock bulkheads, and the air lock barrel.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 53 Significant dimensions of the air lock are as follows:

Parameter Dimension Inside Diameter of Barrel 9'-6" Barrel Thickness 1/2" Door Opening 6'-8"x3'-6" Door Thickness 3/4" Bulkhead Thickness 1 1/8" Each door is locked by a set of six latch pin assemblies, and is designed to withstand the design pressure from inside the containment. To resist the test pressure, each door is fitted with a set of test clamps. The doors are hinged and both swing into the containment. Each door is fitted with two seals that are located such that the area between seals can be pressurized and tested to 52.0 psig. The doors are mechanically interlocked so that only one door can be opened at a time. The capability exists for bypassing this interlock and equalizing the pressure by use of special tools.

The doors may be operated mechanically.

A sight glass is provided in each door to permit observation of the opposite door.

Separate limit switches are provided to allow remote indication of door position and to alarm in the control room.

c. Piping Penetrations Piping penetrations are divided into two types, high energy and moderate energy.

Moderate energy piping penetrations are used for process pipes in which both the pressure is less than or equal to 275 psi, and the temperature of the process fluid is less than or equal to 200F. High energy piping penetrations are used for that piping in which the pressure or temperature exceeds these values.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 54 High energy piping penetrations (Figure 3.8-22) consist of a section of process pipe with an integrally

-forged flued head, a containment penetration sleeve and, where a pipe whip restraint is not provided, a penetration sliding support inside the containment. The sliding support provides shear restraint while permitting relative motion between the pipe and the support. The annular space between the process pipe and the sleeve is completely filled with fiberglass thermal insulation. The pipe and the flued head, are classified as ASME III, Safety Class 2 (NC), whereas the sleeve is classified as part of the concrete containment, ASME III (CC). The sliding support inside the containment is classified as an ASME Safety Class 2 component support (NF).

Moderate energy piping penetrations (Figure 3.8-23) consist of one or more process pipes, the containment penetration sleeve, and a flat circular end

-plate. The pipe is classified as ASME III Safety Class 2 (NC). The sleeve is classified as ASME III Div. 2 (CC). The sleeve backed by concrete is ASME III (CC); the sleeve not backed by concrete is ASME III (Class 2). The end

-plate material is classified as ASME Class

2. Both kinds of assemblies, however, were analyzed to meet the intent of Class MC criteria (NE) through the use of a simplified, but conservative, approach described in Subsection 3.8.2.4. d. Electrical Penetrations Electrical penetrations (Figure 3.8-24) consist of a stainless steel header plate (bulkhead) with an attached terminal box, electrical modules which are clamped to the header plate, and a carbon steel weld ring which is welded to the header plate and to the sleeve. The metallic pressure resisting parts, the sleeve, stainless steel header plate and carbon steel weld ring were designed as ASME III Safety Class MC components (NE); that portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). Double silicone and EPR O-rings provide a seal with a cavity for leakage monitoring between the header plate and the modules. The header plate is provided with a hole on the outside of the containment to allow for pressurization of the penetration assembly for leakage monitoring.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 55 e. Instrumentation Penetrations Instrumentation penetrations are of two types, electrical and fluid. The electrical type is similar in construction to the other electrical penetrations, and the discussion in Subsection 3.8.2.1d, as well as material appearing elsewhere in Subsection

3.8.2 which

pertains to electrical penetrations, is applicable to these penetrations.

The second type of instrumentation penetrations, the fluid type, is similar in construction to the moderate energy piping penetrations. Consequently, the discussion in Subsection 3.8.2.1c on moderate energy piping penetration and other material in Subsection

3.8.2 relating

to these penetrations are also applicable to this type of instrumentation penetration.

f. Fuel Transfer Tube Assembly The fuel transfer tube assembly (Figure 3.8-25) consists of the fuel transfer tube, the penetration sleeve, the fixed saddle on the reactor side, and the sliding saddle in the Fuel Storage Building. The fuel transfer tube and its flanges were designed as ASME III Class 2 components (NC). The quick closure hatch on the refueling canal side of the tube was designed and fabricated to ASME Section III Class MC component (NE) requirements. That portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). The remaining pieces of the assembly were designed as ASME III Component Supports (NF).
g. Ventilation Penetration Assemblies There are two types of ventilation penetrations, the containment air purge penetrations and the containment online purge penetrations. The containment air purge penetrations (Figure 3.8-26) each consist of a pipe sleeve (a rolled and welded pipe section, 36" O.D. by 1/2" wall thickness) which is flanged at each end with 36" weld neck flanges and, attached to these flanges, is a butterfly valve inside containment and a testable blind flange outside containment. Together with the pipe, the blind flanges form a part of the containment pressure boundary during plant Modes 1, 2, 3, and 4. During Modes 5 and 6, the blind flanges are replaced by spool pieces to configure the CAP system to perform its heating and ventilation functions. The valves are 36" diameter butterfly valves with fail

-safe pneumatic operators. The weld between the pipe and the containment liner is equipped with a leak chase for pressure testing.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 56 The containment online purge penetrations each consist of a pipe sleeve (a rolled and welded pipe section, 8" O.D. by 1/2" wall thickness). A short section of pipe with a nipple is welded to the sleeve on the outside of the containment, and a 3/4" valve and test connection is attached to it. The ends of this resulting assembly are welded to 8" weld neck flanges which are through

-bolted to the inner and outer isolation valves. These valves are 8" diameter butterfly valves having fail

-safe pneumatic operators. The weld between the pipe sleeve and the containment liner

is equipped with a leak chase for pressure testing. Since the details, other than size, of these penetrations are essentially the same as those of the containment air purge penetrations, an additional figure is not included.

The portion of the penetration sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). That portion of the sleeve extending beyond the containment was designed to meet the requirements of ASME Section III, Division 1, Subsection NE, in addition to the fabrication and nondestructive examination requirements of Division

2. The valves and flanges were designed as ASME III Code Class 2 components (NC) and are further described in Subsection

3.9.3. SEABROOK

STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 57 3.8.2.2 Applicable Codes, Standards, and Specifications The design, materials, fabrication and inspection requirements for the major containment penetrations conform to, but are not necessarily limited to, the applicable sections of the following codes and specifications which are used to establish or implement design bases and methods, analytical techniques, material properties and quality control provisions.

Dates and revisions given for the listed codes are the earliest version that was used. Subsequent issues were incorporated into the design where practicable, or where the new issue directly affected the safety of the structure.

Code or Specification Title ASME Boiler & Pressure Vessel Code Section II - Material Specification Section III, Division 1, Subsection NA, General RequirementsSection III, Division 1, Subsection NB, Class 1 ComponentsSection III, Division 1, Subsection NC, Class 2 ComponentsSection III, Division 1, Subsection NE, Class MC ComponentsSection III, Division 1, Subsection NF, Component Supports Section III, Divisi on 2 Code for Concrete Reactor Vessels and Containments 1975 Edition including Winter 1976 Addendum for Containment Liner Section IX - Welding Qualifications

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 58 Applicable Codes Dates for ASME B&PV Code, except Section III, Division 2, Components

1971 Edition, through Summer 1973 Addenda (Equipment Hatch) 1974 Edition, no Addenda (Moderate Energy Piping Penetrations) 1974 Edition, through Winter 1975 Addenda (Personnel Air Lock in Equipment Hatch) 1974 Edition, including Winter 1976 Addendum (Electrical Penetrations) 1974 Edition, no Addenda (Breech Type Personnel Air Lock) 1971 Edition through Summer 1973 Addenda (Fuel Transfer Tube Assembly) 1974 Edition, including Summer 1975 Addendum (High Energy Piping Penetrations) 1971 Edition, including Winter 1973 Addendum (Ventilation Penetration Assemblies)

Code or Specification Title U.S. Department of Labor Occupational Safety and Health Administration Standards, October 1975 Edition UBC International Conference of Building Officials, Uniform Building Code, 1973 Edition AISC Specification for the Design Fabrication and Erection of Structural Steel for Buildings, 1969 Edition, (including Supplements 1, 2 and

3) ANSI B16.5

-1968 Steel Pipe Flanges, Flanged Valves, and Fittings ANSI N45.2

-1974 Quality Assurance Program Requirements for Nuclear Power Plants ANSI N45.4

-1972 Leakage Rate Testing of Containment Structures for Nuclear Reactors IEEE STD-317-1972 Electrical Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 59 Code or Specification Title IEEE STD-323-1974 Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations IEEE STD-344-1971 Guide for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations SP-44 Manufacturer's Standardization Society Steel Pipe Line Flanges, 1975 Edition 10 CFR 50, App. J Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors NRC Regulatory Guides Title 1.57 Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components (Rev. 0, 6/73) 1.63 Electrical Penetration Assemblies in Containment Structures for Water Cooled Nuclear Power Plants (Rev. 2, 7/78) 1.84 Code Case Acceptability, ASME Section III Design and Fabrication (Rev. 15, 5/79) 1.85 Code Case Acceptability, ASME Section III Materials (Rev. 15, 5/79) 1.163 Performance

-Based Containment Leak

-Test Program (Rev. 0, 9/95)

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 60 The below-listed UE&C design and construction specifications applicable to the containment penetrations were prepared in accordance with applicable codes, quality control requirements and NRC Regulatory Guides.

UE&C Specifications Title 9763.006-QAS-1 Quality Assurance Administrative and System Requirements 9763.006-QAS-3 Quality Assurance Administrative and Syste m

Requirements For Safety

-Related Electrical Equipment 9763.006-MPS-1 Material and Processing Requirements for Nuclear Power Plant Components 9763.006-NSS-0185 Fuel Transfer Tube 9763.006-WS-1-NE Requirements for Welding and Non-Destructive Examination for Nuclear Pressure Class MC Components 9763.006-SD-15-2 Seismic Requirements for Equipment Hatch and Personnel Air Lock 9763.006-SD-118-1 Seismic Requirements for Electrical Penetrations 9763.006-15-1 Containment Liner 9763.006-15-2 Containment Equipment Hatch and Personnel Locks SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 61 3.8.2.3 Loads and Load Combinations The containment penetrations were designed to withstand all credible conditions of loading, including preoperational test loads, normal startup, operational and shutdown loads, severe environmental loads, extreme environmental loads, and abnormal loads. Design limits and load combinations are in accordance with Article NE

-3000 of Division I and Regulatory Guide 1.57. The penetrations were evaluated for several combinations of loads to assure that the response of the components would remain within the limits prescribed in Subsection 3.8.2.5. In the paragraphs that follow, these loads and load combinations are grouped according to the type of penetration.

a. Equipment Hatch and Personnel Air Lock
1. Design Loads (a) Preoperational Test Loads These are loads which are applied during the initial and any subsequent structural integrity or leak rate testing of the containment.

(1) Test Pressure (P t) The containment and components listed in Subsection 3.8.2.1 are pressurized to 115 percent of the design pressure to test their structural integrity, i.e., the test pressure is 60 psig with design pressure being 52 psig. (2) Test Temperature (T t) The maximum and minimum temperatures inside th e containment during the test are 100F and 50F, respectively. The temperature on the outside of the containment, considered to be the ambient temperature, is 0F, the minimum temperature at the site. Any thermal loads occurring during test conditions are induced by the gradient between the containment and ambient temperatures.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 62 (b) Normal Startup, Operational, and Shutdown Loads Normal loads encountered during normal plant startup, operation and shutdown include the following:

(1) Dead Load (D) This includes the weight of the component and appurtenances.

(2) Live Loads (L) The live load on the raised floors of the personnel locks, representing personnel and equipment that may be moved through there, is taken as 100 pounds per square foot.

There are no other live loads.

(3) Operational Thermal Loads (T o) The normal containment operating temperatures range from 50 to 120F. Outside temperatures vary between 0F and 90 F. (4) Pressure Variation (P v) Differential pressure loads result from pressure variation either inside the containment or in the Containment Enclosure Building. This pressure variation is produced either by atmospheric fluctuations or by HVAC equipment.

The internal pressure varies between

-3.5 and +1.5 psig. (c) Severe Environmental Loads These are loads that would result from external conditions which could infrequently be encountered during the plant life. The following loads are considered in this category:

(1) Wind Load (W) There is no wind load considered due to the presence of the containment enclosure.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 63 (2) Operating Basis Earthquake (E o) These are the loads generated by the Operating Basis Earthquake, which is the earthquake that could reasonably be expected to occur at the plant site during the operating life of the plant.

The applied accelerations, for static analysis, which are valid for rigid components (f > 33 Hz) are 0.580g for horizontal motion and 0.528g for vertical motion; damping is 2 percent for both accelerations. These accelerations represent the peak response of the containment cylinder at the elevation of the equipment hatch, the highest penetration in the containment. Only the actual dead load is considered in evaluating the seismic response forces.

(d) Extreme Environmental Loads Extreme environmental loads are those loads which result from postulated events which are credible but highly improbable. The following loads are included in this category:

(1) Safe Shutdown Earthquake (E s) These are the loads generated by the Safe Shutdown Earthquake, which is the maximum potential earthquake that could occur in the vicinity of the site, based on geological and historical investigations. The applied accelerations, for static analysis, which are valid for rigid components (f > 33 Hz) are 0.861g for horizontal motion and 0.748g for vertical motion; damping is 3 percent for both accelerations. As with the operating basis earthquake, these accelerations also represent the peak response of the containment cylinder at the elevation of the equipment hatch. Only the actual dead load is considered in evaluating the seismic response forces.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 64 (2) Tornado Loads (W t) Due to the presence of the Containment Enclosure Building and other appurtenant structures, such as the pipe chases and the Fuel Storage Building, wind pressure and pressure variation are not considered in the design of containment penetrations.

Furthermore, due to various shield walls the equipment hatch and personnel air lock are protected from impact by tornado-generated missiles. All of the other penetrations (piping, ventilation, etc.), emerge from the containment inside the various appurtenant structures and, consequently, are also protected from impact by tornado

-generated missiles. (e) Abnormal Loads Abnormal loads are those loads generated by postulated high energy pipe ruptures, particularly a rupture in the Reactor Cooling System resulting in a loss

-of-coolant accident (LOCA).

(1) Accident Pressure (P a) The components were designed to withstand an internal containment accident pressure of 52.0 psig. See Subsection 6.2.1 for the development of this pressure.

(2) Accident Temperature (T a) The peak accident temperature of the liner plate is 268F. However, a maximum temperature of 271F was considered for the design of the components. See Figure 3.8-9, Figure 3.8-10 and Figure 3.8-11 for plots of the transient containment liner temperature response, as well as the transient containment pressure response.

(3) Internal Missile Loads Internal missile loads, as described in Section 3.5, are prevented by shields, as required, which confine the missiles.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 65 (f) Environmental Loads These are environmental conditions which must be considered in regard to their effect on the durability of materials.

(1) Normal Relative Humidity Under normal operating conditions, the relative humidity, both inside and outside the containment, is expected to vary between 5 percent and 100 percent. (2) Accident Environmental Conditions Following a loss

-of-coolant accident, the following conditions will exist inside the containment. a Relative humidity of 100 percent b Spray solution of the following:

- Boron (calculated as boric acid),

(minimum/maximum), 0.23/0.25 percent by weight

- pH (minimum/maximum), 9.0/9.6

- Sodium Hydroxide (minimum/maximum),

0.45/0.54 percent by weight

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 66 (3) Radiation Time-integrated doses for personnel air locks and equipment hatch are as follows:

Personnel Air Locks and Equipment Hatch Accident Doses Inside Containment Outside Containment (1 year post accident)

Gamma TID 1.4x10 7 rads 6.5x10 4 rads Beta TID 4.95x10 7 rads ---- Normal Doses (40 years period)

Gamma TID 2x10 7 rads 5.3x10 3 rads Beta TID Negligible Negligible (g) Fatigue Requirements The personnel air locks and equipment hatch were designed to withstand the following conditions

- 120 cycles of plant startup and shutdown

- 400 OBE cycles

- 100 SSE cycles

- 1 accident cycle (LOCA)

- 160 pressure test cycles

2. Load Combinations The load combinations for the equipment hatch and personnel air locks are specified in Table 3.8-6.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 67 b. High Energy Piping Penetrations

1. Loads Design loads applied to high energy piping penetrations were based on selected percentages of plastic load capacities of the attached piping.

Seismic loads were based on the actual seismic response of the containment and piping at the elevation of the penetrations. When actual piping loads were available, having been derived from analysis of the piping systems, they were compared with the design loads; whenever a final load exceeded its design value, a re

-analysis was made to validate the final piping loads. Loads were divided into two categories, Design (which includes normal, upset and emergency conditions) and Faulted, based on operating conditions. Magnitudes of individual design loads are given below. (a) Axial Load (N) For Design Condition, axial load varied from 5 percent to 80 percent of pipe plastic axial load. For the Faulted Condition, 65 percent to 100 percent was used.

(b) Axial Load Due to SSE Conditions (N s) This load was based on the mass of the flued head, and the portion of the sleeve up to the containment wall (outside) and the seismic acceleration at the elevation of the penetration.

(c) Shear Load (V 1) This is the shear applied to the flued head. For the Design condition shear load varied from 5 percent 80 percent of the pipe plastic shear load. For the Faulted condition, 60 percent to

100 percent was used.

(d) Shear Load (V 2) This is the shear in the process pipe at the location of the sliding support. Load V 2 was taken as 10 percent of the pipe elastic shear for the design condition and 100 percent for the faulted condition.

It was determined interactively with the bending moment (M).

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 68 (e) Seismic (SSE) Shear at Flued Head (V s) Load V s is based on the mass of the flued head and the portion of the sleeve up to the outside of the containment wall and the seismic acceleration at the elevation of the penetration.

(f) Pressure Inside the Cavity (P 1) The design pressure inside the cavity was taken as 56 psi, which conservatively represents the containment design pressure (52 psi). (g) Pipe Pressure (P 2) This is the internal design pressure in the process pipe.

(h) Torsion (T) This was taken as 50 percent and 100 percent of the pipe plastic torsion load for the Design and Faulted loading conditions, respectively.

(i) Moment Loads (M 1 and M 2) These moments were considered equal and were taken as percentages of the value of moment load calculated as per the ASME B&PV Code,Section III, Division 1, Subsection NC-3652.1 and the Standard Review Plan 3.6.2, Branch Technical Position MEB 3

-1, Paragraph B(1) (b)(1)(e). For both the Design condition and the Faulted condition, the percentages taken ranged from 60 percent to 100 percent. 2. Load Combinations All loads were co nsidered simultaneously and were grouped into two categories. Design (which includes normal, upset and emergency conditions) and Faulted.

3. Environmental Conditions (a) The normal temperature of the air within the containment enclosure outside the containment structure surrounding the penetration is 104F maximum.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 69 (b) The temperature of the air inside the containment structure is 120F maximum (during normal and upset operating conditions).

(c) Loss of heat from the penetration to the air surrounding the penetration both outside and inside the containment structure is by natural convection.

(d) Process fluid temperatures were considered and vary from penetration to penetration.

(e) No effects of radiation were considered for high energy piping penetrations.

c. Moderate Energy Piping Penetrations
1. Loads Loads applied to the penetration include actual piping loads, as determined from the piping analysis, and other applicable containment design loads.

These loads are in compliance with the intent of Regulatory Guide 1.57 and are described below.

(a) Normal Axial Load (N) This is the axial load imposed by the pipe onto the penetration.

(b) Shear Load (V) This is the shear load imposed by the pipe onto the penetration.

(c) Torsional Load (T) This is the torsional load imposed by the pipe onto the penetration.

(d) Bending Moment (M) This is the bending moment imposed by the pipe onto the penetration.

(e) Pressure Inside the Cavity (P) The design pressure inside the cavity was taken as 56 psig, which conservatively represents the containment design pressure (52 psig).

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 70 (f) Process Pipe Pressure Load (P 1) This is the pressure inside the process pipe.

(g) Deadweight (D) The deadweight of the penetrations and attachments is considered negligible.

(h) Temperature Load (T) This is the load produced by thermal expansion of the attached piping. (i) Operating Basis Earthquake Load (OBE) These are the forces on the penetration from the attached piping caused by the operating basis earthquake acting on the pipe.

(j) Anchor Displacement Loads For the OBE (SAD OBE) These are the forces produced by differential seismic displacements between the penetration and the piping anchors due to the OBE.

(k) Thrust Load (TH) This is the load applied along the axis of the penetration.

(l) Safe Shutdown Earthquake Load (SSE) These are the forces on the penetration from the attached piping caused by the safe shutdown earthquake acting on the pipe.

(m) Anchor Displacement Loads for the SSE (SAD SSE) These are the forces produced from the attached piping by differential seismic displacements between the penetration and the piping anchors due to the SSE.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 71 (n) Pressure and Thermal Anchor Displacement Load (PAD + TAD)

These are the forces on the penetration caused by differential displacement between the penetration and the piping anchors due to pressure and thermal loads.

(o) Pipe Rupture Load (PR) This is the load on the penetration caused by the rupture of an attached pipe.

2. Load Combinations Three design loading conditions were investigated: the Design/Normal Condition (Service Level A), the Upset Condition (Service Level B), and the Emergency/Faulted Condition (Service Level C/D). Load combinations used for each of these design loading conditions are presented in Table 3.8-7. d. Electrical Penetration s 1. Loads (a) Dead Load (D) The weight of the penetration and its attachments were considered.

(b) Test Pressure (P t) The maximum test pressure inside the containment was 60.0 psig.

This is the largest pressure to which the penetrations are subjected and, to simplify the calculations, this is the only pressure that is used for analysis.

(c) Accident Pressure (P a) The maximum containment pressure occurring during a design basis event is 52 psig. (d) Test Temperature (T t) The temperature inside the containment during the pressure test will vary between 50F and 100 F.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 72 (e) Operational Temperature (T o) Under normal operation the temperature inside the containment will vary between 50F and 120 F. (f) Accident Temperature (T a) A peak accident temperature of 271F was utilized for the design of components. This is a design temperature which is higher than the actual temperature to which the penetrations are subjected, and, to simplify the calculation, all material properties are based on this temperature.

See Figure 3.8-27 for a plot of the actual transient temperature response of the electrical penetration seal.

(g) Operating Basis Earthquake (E) A static equivalent load factor of 4.0 was used for both horizontal and vertical earthquakes. This value exceeded all acceleration values calculated in the response spectra at the elevation of the penetrations.

(h) Safe Shutdown Earthquake (E ss) For the SSE, a static equivalent load factor of 4.0 was also used for both horizontal and vertical earthquakes. This value exceeded all acceleration values calculated in the response spectra at the elevation of the penetrations.

(i) External Pressure (P e) Under normal operation, the containment internal pressure varies between +1.5 psig and -3.5 psig. Thus, the maximum "external" pressure during normal operation was 3.5 psig. 2. Environmental Conditions (a) Outside Temperature The electrical penetrations were designed for temperatures outside the containment which vary between 40F and 104F. Following a DBA, the maximum temperature is 130 F.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 73 (b) Atmospheric Pressure Pressure on the outside of the containment varies between +0.006 psig and -0.005 psig during normal operation and between 0 psig and -0.011 psig during the containment test. Following a DBE, the minimum pressure is

-0.013 psig. (c) Relative Humidity The maximum relative humidity inside the containment is

90 percent and outside the containment is 95 percent. Following a DBE, the relative humidity inside the containment is 100 percent. (d) Radiation The normal radiation rate is 50 millirads per hour. During the containment test, the radiation rate is 15 millirads per hour. The total cumulative radiation, considering 40 years of normal operation, a DBE and the post DBE period, is 1.3x10 8 Rads. (e) Chemical Sprays Chemical sprays include the following:

- Boric Acid 0.21 (by weight)

- pH (min/max) 8.5/10.5 - Sodium Hydroxide Solution 0.42 (by weight)

3. Loading Combinations Loading combinations are given in Table 3.8-8.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 74 e. Fuel Transfer Tube Assembly

1. Design Loads (a) Preoperational Test Loads (1) Test Pressure (P t) The test pressure is 60 psig. (2) Test Temperature (T t) The temperature range inside the containment is 50F to 100F, and the minimum temperature at the site is 0 F. (b) Normal Loads (1) Dead Load (D) This includes the weight of the fuel transfer tube and its support system.

(2) Live Load (L) This includes the weight of the fuel assembly and carriage and the weight of water in the fuel transfer tube.

(3) Pressure Variation (P v) The containment internal pressure varies between

-3.5 and +1.5 psig. (4) Hydrostatic Pressure (H) The maximum depth of water, above the tube, during refueling is 34'

-41/4". (c) Severe Environmental Loads (1) Wind Loads (W) There is no wind load due to the presence of the Fuel Storage Building.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 75 (2) Operating Basis Earthquake (E o) The accelerations produced by the operating basis earthquake, which are valid for rigid components (f >33 Hertz), are 0.45g for N

-S motion, 0.79g for E

-W motion, and 0.42g for vertical motion. The OBE seismic loads also included the forces produced by differential seismic displacement between the Containment Building and the Fuel Storage Building.

(d) Extreme Environmental Loads (1) Tornado Loads (W t) Tornado loads are not considered due to the presence of the Fuel Storage Building.

(2) Safe Shutdown Earthquake (E s) The accelerations produced by the safe shutdown earthquake, which are valid for rigid components (f >33 Hertz), are 0.75g for N

-S motion, 1.12g for E

-W motion, and 0.61g for vertical motion.

The SSE seismic loads also included the forces produced by differential seismic displacement between the Containment Building and the Fuel Storage Building.

(e) Abnormal Loads (1) Accident Pressure (P a) The containment internal accident pressure of 52.0 psig was considered in the design of the fuel transfer tube assembly. (2) Accident Temperature (T a) A peak accident temperature of 271F was considered for the design of the components.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 76 (f) Fatigue Requirements The fuel transfer tube was designed to withstand the following conditions:

- 400 OBE cycles

- 1 accident cycle (LOCA)

- 160 pressure test cycles

- 1000 temperature cycles

2. Load Combinations The load combinations are given in Table 3.8-9. f. Ventilation Penetration Assemblies
1. Design Loads (a) Deadweight (D) The deadweights of the sleeve, valves and attached piping, as applicable, were included.

(b) Operating Basis Earthquake (OBE) The inertia forces produced by the operating basis earthquake were included. (c) Safe Shutdown Earthquake (SSE) The inertia forces produced by the safe shutdown earthquake were included. (d) Normal Containment Pressure (P n) The normal containment pressure varies between

-3.5 and +1.5 psig.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 77 (e) Accident Pressure (P f) The accident pressure on the containment due to either a LOCA or a steam line break, whichever is higher, is used. The LOCA pressure governs and is 52.0 psig for design purposes.

(f) Test Pressure (P t) The containment test pressure is 60.0 psig. 2. Load Combinations (a) Normal and Upset D + OBE + P n (b) Faulted D + SSE + P f (c) Test D + P t 3.8.2.4 Design and Analysis Procedures The portions of the containment classified as steel containment were designed and analyzed using procedures described below. The components were designed to safely withstand the load combinations defined in Subsection 3.8.2.3. a. Personnel Air Lock The personnel air lock was designed and analyzed in accordance with Subsection NE of the ASME Boiler and Pressure Vessel Code,Section III, Division 1. The air lock door was designed for an internal pressure of 52 psi combined with an equivalent pressure due to the OBE (0.40 psi) and the SSE (0.60 psi). The total pressure was applied to the complete cover. Design was governed by the SSE load in the Faulted Condition; hand computations were used.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 78 That portion of the air lock barrel not backed by concrete was analyzed by hand computations for the most critical load combination in the Faulted Condition.

The internal pressure and an additional pressure representing the effect of horizontal seismic loading result in a total pressure of 52.54 psi on the vertical projection of the shell. The deadweight and vertical seismic loading result in a vertical load of 20,277 lbs.

Stresses and strains within the barrel due to displacements imposed by the concrete were determined using a finite element model and the computer program ANSYS described in Subsection 3.8.2.4i. The displacements of the concrete were determined in the three

-dimensional finite element analyses described in Subsection 3.8.1.4. Both elastic and elastic

-plastic behaviors were considered.

Using hand computations, it was also determined that the natural frequencies associated with axial vibration of the air lock door and of the barrel as well as transverse vibration of the barrel are all much greater than 33 Hertz. Therefore, the personnel air lock was considered rigid for purposes of seismic analysis. The natural frequency for vertical vibration of the door in the open position, however, is less than 33 Hertz. This condition is not under the jurisdiction of Subsection NE but was considered for the design of the hinge plates and pins.

b. Equipment Hatch The equipment hatch was designed and analyzed in accordance with Subsection NE of the ASME Boiler and Pressure Vessel Code,Section III, Division 1. The equipment hatch cover was designed for the effects of accident pressure combined with the SSE, with the total applied pressure being 52.40 psi. Since live load and dead load are negligible, this combination represents the Faulted Condition with SSE, load combination 5 in Table 3.8-6. The flange, bolts, and other components were also designed for the above indicated load combination as well as for external pressure, bolt preload, and other loads produced by gasket seating and seal testing. Hand computations were used. Rotation of the flange was calculated to verify seal integrity, using an axisymmetric finite element model and the computer program AX2 (described in Subsection 3.8.2.4i). The critical loads for this analysis are bolt preload, with pressure on the convex side of the head, and bolt preload acting alone. In both cases, the integrity of the seal was not violated.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 79 The design of that portion of the barrel not backed by concrete was based on analysis for accident pressure and for test pressure. Dead load and seismic load are negligible and were omitted. An axisymmetric shell model was analyzed using the computer program ANSYS. In these analyses, two sets of boundary conditions were used, one in which the flange was considered rigidly attached to the barrel and another in which the far end of the barrel was free. In this manner, the actual restraint was bounded.

A similar finite element model was used to determine the minimum thickness required for the end portion of the barrel, which is backed by concrete. Results at the concrete

-to-metal junction verified the adequacy of a thickness of 1 inch.

Buckling of the head, sleeve, and barrel was checked, using the procedures of Subarticle NE

-3133 of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NE, and found not to occur. Fatigue in the barrel was also investigated, and the design was found to be satisfactory.

To verify the design of the entire equipment hatch assembly, displacements imposed by the concrete onto the barrel were considered in an axisymmetric shell model which can accept nonaxisymmetric loading. The entire assembly, cover, flange and barrel, were modeled using ANSYS. Displacements were imposed at the intersection of the barrel with the containment liner, and accident pressure and temperature, dead load, live load, and seismic load were also included. Note that the personnel air lock was omitted from this model because simplified analyses of that region showed it to be similar in stiffness to the equipment hatch. There were four sets of imposed displacements based on four different cracking patterns in the concrete. For all load combinations, strains were added on an elastic basis and found to be below allowable values.

The air lock mounting sleeve, including its reinforcement, was analyzed by hand computation. Loads considered included accident pressure, dead load, live load, seismic load, and steady

-state temperature. Stress analysis of the sleeve at the sleeve-collar boundary was performed using the AX2 axisymmetric shell program, considering only test pressure.

Transverse natural frequencies of the hatch, sleeve, and air lock were all found by hand computations to be greater than 33 Hertz. Using axisymmetric shell elements and the computer program ANSYS, the axisymmetric modes of vibration of the hatch were also found to be greater than 33 Hertz. Therefore, the equipment hatch assembly was considered rigid for purposes of seismic analysis.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 80 c. Personnel Air Lock in Equipment Hatch The personnel air lock in the equipment hatch was designed and analyzed in accordance with Subsection NE of the ASME Boiler and Pressure Vessel Code,Section III, Division

1. This air lock was analyzed for accident pressure and SSE seismic loads. Overall structural response was calculated using the computer program STARDYNE (described in Subsection 3.8.1.4) and a three

-dimensional model. Localized stresses at attachments, such as latches, pins and bearings, were determined using hand computations. In the analysis of the latches, an external pressure of

(-)3.5 psig was applied in conjunction with the seismic loads.

The personnel air lock floor was designed for live load and dead load, and the test clamps, bolts, and plates were designed for the test pressure, all using hand computations. The open door seismic condition was also checked, considering the door in this position to be seismically flexible.

However, for all other design conditions the natural frequencies of the air lock assembly were found to be greater that 33 Hertz, and the assembly was considered rigid for purposes of seismic analysis. These calculations for natural frequency were made using the same finite element model that was used to determine overall response of the air lock.

d. High Energy Piping Penetrations High energy piping penetrations were designed and analyzed to meet the intent of Subsection NE of the ASME Boiler and Pressure Vessel Code,Section III, Division 1. The design insured that the sleeve have a greater load carrying capacity than the process pipes.

The analysis procedure utilized three phases.

The first phase consisted of a heat transfer analysis to determine the temperature in the flued head, the sleeve and at the sleeve/containment wall interface. This was accomplished through the use of a finite element model and the computer program MARC

-CDC (described in Subsection 3.8.1.4). After establishing that concrete temperatures complied with the limit of ASME Section II, Division 2, Subsection CC-3340(b), a finite element model for stress analyses was established. This second phase consisted of a stress analysis utilizing unit load cases representing axial (N), shear (V), torsion (T), and moment (M) loadings.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 81 Two additional load cases were run, using the containment design pressure and the actual thermal gradient, as determined in the heat transfer analysis. Total stresses for the design loads presented in Subsection 3.8.2.3 were determined by multiplying stresses from the unit load cases by the appropriate load ratio and combining the results with those from the pressure and thermal gradient runs.

The third phase consisted of using stress results from phase two to determine stress intensities at various locations in the sleeve and flued head and comparing these intensities to allowable values.

Computer programs used for stress analysis included MARC

-CDC, WILSON 1 and WILSON 2 (all described in Subsection 3.8.1.4), and ANSYS (described in Subsection 3.8.2.4i). The temperature gradient which was used in the stress analyses was determined using the computer program TAPAS (described in Subsection 3.8.1.4); thermal boundary conditions were determined in the MARC-CDC heat transfer analysis of phase one.

e. Moderate Energy Piping Penetrations Moderate energy piping penetrations were designed and analyzed to meet the intent of Subsection NE of the ASME Boiler and Pressure Vessel Code,Section III, Division
1. The design insured that the sleeves have a greater load carrying capacity than the process pipes.

The analysis procedure consisted of subjecting the penetration assembly to actual pipe loads and other applicable containment loadings and determining stresses at four critical locations, which are the junctions of the sleeve with the containment wall and the end plate and the junctions of the end plate with the sleeve and the process pipe. Weld stresses at the pipe/end plate interface weld were also determined.

All stresses were compared with appropriate allowable values to determine the adequacy of the penetrations.

All computations were performed by hand using standard formulas and techniques taken from established references and/or by the finite element program ANSYS. f. Electrical Penetrations Electrical penetrations were designed and analyzed in accordance with Subsection NE of the ASME Boiler and Pressure Vessel Code,Section III, Division 1.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 82 A typical penetration was analyzed for dead load, test pressure (being the larger of maximum LOCA pressure and test pressure) and safe shutdown earthquake loads. The temperature was taken as 375F for all loading conditions.

It was considered unnecessary to do a thermal analysis since the maximum temperature the assembly will be exposed to is only 370F. Fatigue analysis was also unnecessary since it is expected that only a few cycles of loading will be experienced in the 40

-year life of the plant.

Using appropriate hand computations, the below

-listed penetration components were analyzed. Items 1 through 9 comprise the pressure boundary and Items 1 through 8 code (Division

1) items. Items 10 and 11 are accessories which were also analyzed.

Item 1 - Installation weld which attaches the penetration assembly to the nozzle Item 2 - 1/2-13 stud which holds the single clamp in place Ite m 3 - 5/8-11 stud which holds the center clamp in place Item 4 - Module - The module transmits the pressure applied over its projected area to the clamps.

Item 5 - Monitoring plate Item 6 - Attaching weld (Monitor plate to flange extension)

Item 7 - Center clamp which receives loading from three modules. Each module contributed 1/3 of the load applied to it.

Item 8 - Single clamp. Each module contributed 1/3 the load applied to it. Item 9 - Epoxy sealant material in the penetration module Item 10 - Penetration terminal box (outboard)

Item 11 - Penetration terminal box (inboard).

To further verify the integrity of the penetration, it is pressure tested as described in Subsection 3.8.2.7.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 83 g. Fuel Transfer Tube Assembly The fuel transfer tube and its support system were designed and analyzed in accordance with Subsection NC of the ASME Boiler and Pressure Vessel Code,Section III, Division

1. The fuel transfer tube assembly was analyzed for the various load combinations as listed in Table 3.8-9. Overall structural response was calculated using the computer program STARDYNE and a three

-dimensional finite element model. This model was also used to determine the natural frequencies of the assembly, the lowest of which was 34.14 Hertz. Therefore, the fuel transfer tube assembly was considered rigid for purposes of seismic analysis.

Buckling of the tube was also checked, using the procedures of Subarticle

NC-3133 of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NC, and found not to occur.

In the load combinations which include seismic loads, the relative displacements due to differential movements between the Containment Building and the Fuel Storage Building were imposed at the sliding support.

h. Ventilation Penetration Assemblies Ventilation penetrations were designed and analyzed to meet the intent of Subsection NE of the ASME Boiler and Pressure Vessel Code,Section III, Division 1. The design was based on system requirements, and subsequent analysis indicated that stress levels were well below an acceptable limit.

The analysis consisted of conservative static analysis procedures (hand computations) to determine stress levels at the interface with the outside of the containment wall, the critical region of the assemblies. Both primary and secondary stresses were calculated.

i. Descriptions of Computer Programs Utilized in the Design and Analysis The computer programs used in the design and analysis of the steel components (Section III, Division 1 items) of the concrete containment which resist pressure and are not backed by concrete, are described briefly in this subsection. Programs listed below are only those programs which have not been previously described in Subsection 3.8.1.4g. A summary of the comparisons of results used to validate them is given in Appendix 3F. The program classifications are discussed in Subsection 3.8.1.4g.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 84 1. ANSYS: Engineering Analysis System, by Swanson Analysis System, Inc. (Houston, Pa.). Documentation is available from Control Data Corporation. ANSYS provides static and dynamic analysis capabilities, including plasticity, creep and swelling; small and large deflections; steady-state and transient heat transfer; and steady fluid flow.

2. AX2: Axisymmetric Shell Program, by Pittsburgh-Des Moines (PDM) Steel Company. Documentation is available from PDM. AX2 is a computer program for the analysis of single layer, axisymmetric thin shells of revolution. Materials are isotropic and elastic, and loads are static and axisymmetric.

3.8.2.5 Structural Acceptance Criteria The components classified as steel containment were designed to remain within the design limits specified in Subsection NE of the ASME B&PV Code,Section III, Division 1, and NRC Regulatory Guide 1.57. Code boundaries for all penetrations except the personnel air lock are the portion of the sleeve which is backed by concrete and the attachment weld to the liner are Division 2; all remaining portions of the penetrations are Division

1. The division boundaries for the personnel air lock are described in Subsection 3.8.2.6c. Additional acceptance criteria pertaining to specific penetrations are described below.
a. Equipment Hatch and Personnel Air Locks Allowable stresses for the equipment hatch and personnel air locks are given in Table 3.8-10 and Table 3.8-11. b. High Energy Piping Penetrations The high energy piping penetrations were classified as ASME Section III, Division 1, Safety Class 2 components (NC). However, since the stress limits of Subsection NC are not directly applicable to this type of component for the specified categories of operating conditions, the penetrations were analyzed to meet the intent of the requirements of ASME Section III, Division 1, Subsection NB, and the allowable stress intensities were limited to the values given in Subsection NC. These allowables are summarized, in general terms, in Table 3.8-12.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 85 c. Moderate Energy Piping Penetrations The moderate energy piping penetrations were designed to meet the intent of ASME Section III, Division 1, Class MC (NE) criteria for stress limits and load combinations. Stress limits, used to determine the acceptability of the penetrations, are as follows:

Primary Stress Intensity Primary & Secondary Stress Intensity

Membrane Membrane + Bending Design/Normal Conditions 1.0 Smc 1.5 S mc 3 S mi Upset Conditions 1.0 S mc 1.5 S mi 3 S mi Emergency/Faulted Conditions 1.2 S mc 1.0 S y 1.8 S mc 1.5 S y NA d. Electrical Penetrations All stresses are classified in accordance with Table NE

-3217-1 of the ASME Code,Section III, Division 1, Subsection NE, and maximum stresses were found to be within the following limits:

1. The design limit of Subsection NE-6322 was not exceeded when the assembly was subjected to test pressure.
2. The design limits of Subsectio n NE-3131 (a), (b), and (d) were not exceeded when the assembly was subjected to the effects of the load combinations for test, normal, upset and faulted (abnormal/severe environmental) conditions.
3. The design limits of Subsection NE-3131 (c) (1) or (2), as applicable, were not exceeded when the assembly was subjected to the effects of the faulted (abnormal/extreme environmental) load condition.
4. The design limits of Subsection NE-3131.1 were not exceeded when the assembly was subjected to concurrently applied design loadings that produced the greatest potential for shell instability and loadings associated with the vibratory motion of the safe shutdown earthquake.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 86 e. Fuel Transfer Tube Assembly The fuel transfer tube assembly was designed to meet the intent of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NC, for Class 2 components. The maximum stresses and displacements at the bellows were found to be within the following limits:

1. The design limits of Subsection NC-3200, Appendix XIII

-1140 were not exceeded when the assembly was subjected to the effects of the load combinations representing test, normal, refueling and upset conditions.

2. The design limits of Subsection NC-3200, Appendix XIII

-1140 were not exceeded when the assembly was subjected to the effects of the faulted load condition.

3. The fuel transfer tube was analyzed in accordance with Subsection NC-3133 for the cylindrical shell subject to test pressure, accident pressure and seismic loadings produced by the SSE. 4. The deflection capabilities of the bellows were not exceeded when the assembly was subjected to the effects of the load combinations for test, normal, refueling, upset and faulted conditions.
f. Ventilation Penetration Assemblies The ventilation penetrations were designed to meet the intent of ASME Section III, Division 1, Class MC (NE) criteria for stress limits and load combinations. For all of the indicated load conditions the stress limit was 1.0 S
m. 3.8.2.6 Materials, Quality Control and Special Construction Techniques
a. Materials Materials used for components classified as steel containment include steel plate, bolts, pins, seamless pipe, seals, welding filler materials, coatings, and other miscellaneous items. These materials are listed in Table 3.8-11 for the equipment hatch and personnel air locks, and in Table 3.8-13 for the piping, electrical, and instrumentation penetrations, fuel transfer tube and ventilation penetration assemblies. Not listed in these tables is the protective coating system which is identical to the system described in Subsection 3.8.1.6. These coatings are applied to all exposed surfaces.

Replaceable silicone seals are used in the equipment hatch and personnel air lock.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 87 b. Special Construction Techniques The only special construction technique employed was the manner in which the personnel air lock was fabricated. To maintain the construction schedule, the barrel, inside bulkhead and outside bulkhead were shipped to the site and assembled and tested in place.

The testing described in Subsection 3.8.2.7 verified that this assembly procedure did not affect the integrity of the lock.

The equipment hatch was also fabricated in one piece to minimize field welding.

Shop and field testing for the hatch are described in Subsection 3.8.2.7. c. Quality Control

1. Equipment Hatch and Personnel Air Locks The contractor certified by application of the appropriate code symbol and completion of the appropriate Data Report in accordance with Article

NE-8000 (Division

1) that the material used complies with the requirements of Article NE

-2000 (Division 1), and that the fabrication, installation, and construction comply with the requirements of Article

NE-4000 (Division 1). Specific quality control requirements are further discussed below.

Quality control procedures were in accordance with Articles NE

-4000 and NE-5000 of Subsection NE of the ASME Code,Section III, Division

1. For all components, including welding and brazing materials covered under this section, the fabricators supplied Certified Materials Test Reports. These reports included chemical analyses, physical tests, mechanical tests, examinations and heat treatment.

Physical tests include a Charpy

-V-Notch toughness test performed in accordance with Article NE

-2300 of Division

1. The test temperature was

(-)10F in the heat

-affected zone and (-)25F in the parent metal for the equipment hatch and personnel air locks.

Longitudinal seams in welded pipe were 100 percent radiographed.

Dimensional standards for individual metallic components were in accordance with Section NE-2700 (Division 1).

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 88 All welding was performed in accordance with NE

-4000 (Division 1). When the base metal temperature was less than 50F, welding was not performed without heating the metal.

When base metal temperature was within the range of 0F to 50F inclusive, the base metal within 3" of the place where welding was to be started was heated to a minimum temperature of 50 F. Fabrication of the equipment hatch was performed in accordance with NE-4000 (Division

1) using the best shop practices relative to edge treatment, alignment, and general workmanship.

Due to its field fabrication, the personnel air lock was stamped as follows:

the center barrel from 8" inside the containment to 1'

-3" outside the containment is NPT Class CC, Division 2, 1975, with the Data Report noting that parts extending beyond the concrete were fabricated in accordance with Division 1, 1974 Edition. The other two end sections were stamped NPT Class 2 or MC Division 1, 1974 Edition.

Fabrication tolerances of linear dimensions, such as length and diameter of hatch and locks, were as listed below unless specified otherwise on drawings. Dimensions less that 6 inches +/-1/32 inch Dimensions 6 inches to 2 feet-6 inches +/-1/16 inch Dimensions over 2 feet-6 inches to 10

+/-1/8 inch Dimensions greater than 10 feet +/-1/4 inch Erection tolerances insured that the equipment hatch and personnel air locks were set plumb, square and level and at their proper elevation and plane. Angular tolerance through the containment wall, as measured from the point of attachment of each air lock or the equipment hatch to the containment liner, was not greater than +/-30 minutes. Angular tolerance of the theoretical point of attachment to the containment liner, as measured from the center of the containment, was +/-4 minutes in the horizontal plane. Vertical tolerance was +/-1 inch. Negative radial deviation from the theoretical vertical centerline did not exceed (-)1/4" at the equipment hatch.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 89 2. Other Penetrations Fabrication criteria for the other penetrations have been discussed in Subsection 3.8.2.1 with the descriptions of the penetrations.

3.8.2.7 Testing and In

-Service Surveillance Requirements

a. Equipment Hatch and Personnel Air Locks The structural testing and in

-service inspection program consists of the following:

1. Strength test of personnel locks
2. Leak test of personnel locks
3. Leak test of leak chase system
4. Leak test of seals
5. Joint inspection
6. Other testing as described in Subsection 3.8.1.7. Procedures are described in the paragraphs which follow.

The strength test is conducted in accordance with NE

-6320 (Division 1); it consists of pressurizing the lock to 60.0 psig and then holding the pressure for 15 minutes with hold

-down bars on the inner door.

Following the strength test, a leak test is conducted in accordance with NE

-6215 (Division 1). During this test, the air pressure is reduced to 52.0 psig and held for 15 minutes. If a pressure drop occurs, soap bubble tests are conducted, repairs made, and the personnel lock retested.

Leak chase systems are leak tested in accordance with NE

-6215 (Division

1) by pressurizing to 52.0 psig and holding for 15 minutes. If a pressure drop occurs, soap bubble tests are conducted, repairs made, and the personnel lock retested.

Seals for personnel lock doors, equipment hatch and at other locations are leak tested by pressurizing the space between the seals to 52 psig and holding for 15 minutes. If a pressure drop occurs, a halogen diode detector test was conducted, repairs made, and the seals retested.

All joints not covered by test channels are accessible for in

-service inspection.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 90 Testing of the entire containment structure, which includes the equipment hatch and personnel air lock, includes the integrated leak rate test, the in

-service leak rate test, the preoperational structural integrity test, and general visual inspection of structurally critical areas. All of these tests are described in either Subsection 3.8.1.7 or in Subsection 6.2.6. b. Other Penetrations The portions of all the other penetrations which form part of the containment pressure boundary are subjected to the test program performed on the completed containment. This program includes the integrated leak rate test, the in

-service leak rate test, the pre

-operational structural integrity test and general visual inspection of structurally critical areas, all of which are described in either Subsection 3.8.1.7 or Subsection 6.2.6. The penetrations also receive certain design verification testing as described below. A typical electrical penetration is pressure tested in order to demonstrate that the module, epoxy and "O" ring seal can withstand the required pressure. A pressure chamber is fitted with the module and the pressure is raised to 6,000 psig without perceptible damage to the "O" rings or the penetration module. Thus a safety factor of 100 can be assumed (since the maximum design pressure is 60 psig). 3.8.3 Concrete and Structural Steel Internal Structures The following section contains the physical description, codes, loads and load combinations, design and analysis procedures, allowable stresses, quality control and testing as they relate to the internal structures of the containment.

3.8.3.1 Description of the Internal Structures The description of the structural configuration, materials, location and arrangement of the internal structures appears on the general arrangement drawings, Figure 1.2-2, Figure 1.2-3, Figure 1.2-4, Figure 1.2-5, and Figure 1.2-6.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 91 The following major internal structures are located in the containment structure. They incorporate no unique or new design or construction features. Also, block or concrete masonry partitions are not utilized in the containment.

a. Reactor Support Syste m The reactor vessel is supported by two distinct support systems, the vertical box support and the lateral ring girder support. The vertical box support transmits vertical loads from the reactor vessel to the primary shield wall; it also transmits lateral loads from the reactor vessel to the lateral ring girder, which then transmits the loads to the primary shield wall. Design and analysis of the vertical box support are described in Subsection 5.4.14. The lateral ring girder support is described below

. The ring girder, reactor vessel lateral support, consists of a set of four curved girders resting on the primary shield wall at Elevation (-)14'-1", constructed of welded plates of ASTM A588, Grade 50 steel. As indicated above, its function is to restrain the lateral movement of the reactor vessel and to transfer the resultant lateral loads to the primary shield wall through normal contact. For details see Figure 3.8-28. b. Steam Generator Support System The steam generators are supported by several steel component supports, including vertical supports, upper lateral supports, and lower lateral supports. The vertical supports transmit loads to the fill mat and the upper and lower lateral supports transmit loads to both the primary and secondary shield walls. See Subsection 5.4.14 for a description of these supports.

c. Reactor Coolant Pump Supports Reactor coolant pump supports, including vertical columns and lateral tension tie bars, are described in Subsection 5.4.14. d. Primary Shield Wall The primary shield wall is a circular reinforced concrete wall, varying in thickness from 4'

-9" to 8'-6", enclosing and supporting the reactor pressure vessel. It is supported on the fill mat slab and extends to the refueling canal bottom elevation. In addition to providing shielding to the interior containment during normal operation or maintenance, the wall protects the reactor vessel from blowdown effects in the event of a rupture in the primary system piping outside the wall.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 92 The reinforcing consists of several layers of hoop and meridional bars with the meridional bars extending either into the fill mat or the continuation of the fill mat down in the cavity in order to provide anchorage for the wall. See Figure 3.8-29. e. Secondary Shield Wall The secondary shield wall is a 4

-foot thick octagon shaped reinforced concrete wall enclosing the reactor coolant piping, steam generators, reactor coolant pumps and their supports. The shield wall has openings to vent the area and to permit access from the outer annular area of the containment structure. These openings are protected by offset reinforced concrete walls which provide radiation shielding and protection from pipe whip. The secondary shield wall is supported on the fill mat slab and the above described portion extends to the underside of the operating floor.

Extending upward from the operating floor are those portions of the secondary shield wall that are referred to as the biological shield walls. They extend 7 feet above the operating floor and are shaped, in plan, like an elongated octahedron. Each one encloses two steam generators and provides radiation shielding.

f. Refueling Canal The refueling canal consists of reinforced concrete walls and floors lined with ASTM A240 Type 304 stainless steel to provide a leak

-tight membrane during refueling operations. The floor extends from the reactor cavity to the fuel transfer penetration and is supported on the primary shield wall and support walls which extend to the fill mat. The walls extend from the base of the canal to the operating floor level. During refueling the new and spent fuel elements are transported through the canal, which is flooded to Elevation 23.5 feet. Tools used during refueling are supported from the canal walls. For a more detailed description of the refueling canal and its use during refueling, see Subsection 9.1.4. g. Pressurizer Region The pressurizer region consists of a shield wall which provides shielding and missile protection. The wall is composed of several precast concrete sections which are placed one atop another. To maintain continuity these sections are bolted together. These sections are provided with lifting lugs and, when unbolted, may be removed to permit access to the pressurizer for maintenance and

in-service inspection.

Pressurizer supports are described in Subsection 5.4.14.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 93 h. Control Rod Drive Mechanism (CRDM) Missile Shield and Cable Bridges The CRDM missile shield has been incorporated into the simplified head assembly. The original missile shield structure has been removed from the Containment. To support the CRDM and Drive Rod Position Indication (DRPI)

System cables that run from the simplified head assembly two hinged cable bridges are provided on the operating deck, one on each side of the refueling cavity. See Figure 1.2-6A. i. Operating Floor Slab The operating floor slab is a reinforced concrete slab, 3'

-0" to 8'-6" thick, located at Elevation +25.0 feet spanning between the crane support columns and the secondary shield walls. There are openings in the floor for access, installation, and removal of the four steam generators and the pressurizer, and access to the refueling canal and internal storage area. Removable reinforced concrete plugs are provided over the reactor coolant pumps and the in-core detector drive. The polar crane rail is located at the extremities of the operating floor concrete slab over its support structure.

j. Fill Mat The internal structures are supported on a 4

-foot thick reinforced concrete fill mat poured over the liner on the structural base mat. This fill mat is not connected to the base mat. Horizontal load transfer is provided by the keying action of the fill mat in the reactor pit, elevator pit, and sump pits. The fill mat extends under the secondary shield wall and the crane support columns to the containment cylindrical wall. An expansion joint is provided between the end of the fill mat and the containment cylindrical wall. The fill mat is thus independent of the primary containment structure walls.

All supports for primary equipment and for other components and equipment located in the annulus area outside the secondary shield wall are supported on the fill mat.

k. Structural Steel Structural steel is provided to support floors in the annulus at the operating level and in other areas of the containment structure where access to components and equipment is required, to provide a means of travel from various quadrants of the containment, and for pipe support at lower levels. Steel grating and concrete slabs are provided for a walking surface, and stairways and an elevator are required for movement between various levels of the annular steel.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 94 l. Crane Support Structure The crane rail is located on a continuous reinforced concrete beam at the operating floor level. The beam is supported on a series of reinforced concrete columns which extend to the fill mat outside the secondary shield wall. The beam is also connected to the biological shield wall through the floor slab at the operating floor level.

This tie reduces the torsional effect on the beam, which is designed to resist the remaining torsional effects. Lateral stability of the columns is provided by reinforced concrete beams and slabs connected to the secondary shield wall at Elevations 0' and 25'. The crane specification requires that the manufacturer's design insures that the crane will remain on the rail during either an OBE or an SSE. This is accomplished by means of kick

-back plates attached to the wheel carriage, which prevent separation between the wheel and the rail. See Figure 3.8-30. Seismic base shears are transmitted to the floor slab by means of a positive connection between the rail and the floor slab, which is part of the supporting structure. The crane rail anchor is designed to transfer all shear forces and overturning moments (resulting from the action of either an OBE or an SSE on the crane) to the concrete supporting system. Details of this connection include clamp plates, which restrain the rail and are welded to the base plate, and anchor bolts, which secure the base plate to the concrete beam. See Figure 3.8-30. m. Neutron Shielding The neutron shielding is a shell structure mounted on the slabs and walls above and around the reactor vessel. It consists of a plastic material encased in steel plate and provides radiation shielding.

A more complete description of the neutron shielding and its effect on structures can be found in Section

6.2. SEABROOK

STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 95 3.8.3.2 Applicable Codes, Standards and Specifications The design, materials, fabrication and inspection of the internal structures conform to, but are not necessarily limited to, the applicable sections of the following codes and specifications which are used to establish or implement design bases and methods, analytical techniques, material properties and quality control provisions. Dates and revisions given for the listed codes are the earliest version that was used. Subsequent issues were incorporated into the design where practicable or where the new issue directly affected the safety of the structure.

Code or Specification Title ACI 211.1-70 Recommended Practice for Selecting Proportions for Normal Weight Concrete ACI 214-65 Recommended Practice for Evaluation of Compression Test Results of Field Concrete ACI 301-72 Specification for Structural Concrete for Building (application sections) ACI 302-69 Recommended Practice for Concrete Floor and Slab Construction ACI 304-73 Recommended Practice for Measuring, Mixing, Transporting and Placing Concrete ACI Committee Report 74-33 Recommended Practice for Hot Weather Concreting ACI Report 306R

-78 Recommended Practice for Cold Weather Concreting ACI 308-71 Recommended Practice for Curing Concrete ACI 309-72 Recommended Practice for Consolidation of Concret e ACI Committee Placing Concrete by Pumping Method ACI Committee Report

72-33 Placing Concrete with Belt Conveyors ACI 311-64 Recommended Practice for Concrete Inspections

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 96 Code or Specification Title ACI 315-65 Manual of Standard Practice for Detailing of Reinforced Concrete Structures ACI 318-71* Building Code Requirements for Reinforced Concrete (with Commentary)

ACI 347-68 Recommended Practice for Concrete Formwork ACI 614-59 Recommended Practice for Measuring, Mixing, and Placing Concrete ACI SP2 (1975 Edition) ACI Manual of Concrete Inspection CRSI Reinforced Concrete

- Manual of Standard Practice, 22nd Edition, first printing, 1976 CRSI Recommended Practice for Placing Reinforcing Bars, 1968 ASME ASME Boiler and Pressure Vessel Code,Section II, Material Specification, Part C

- Welding Rods, Electrodes and Filler Metals (up to and including Winter 1974 Addenda)

ASME ASME Boiler and Pressure Vessel Code,Section III, Division 2 - Code for Concrete Reactor Vessels and Containments (1975 Edition) ASME ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NF (up to and including Winter 1974 Addenda)

ASME ASME Boiler and Pressure Vessel Code,Section V, Nondestructive Examination (up to and including Winter 1974 Addenda) ASME ASME Boiler and Pressure Vessel Code,Section VIII, Pressure Vessels, Part 1 (up to and including Winter 1974 Addenda)

  • Since the internal structures are reinforced and not prestressed, cannot be classified as thin shells or special ductile frames (no formation of plastic hinges allowed from seismic loads) and are not designed as special shear walls, the provisions of ACI 318

-71 code Chapters 18, 19 and Appendix A are not applicable.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 97 Code or Specification Title ASME ASME Boiler and Pressure Vessel Code,Section IX, Welding and Brazing Qualifications (up to and including Winter 1974 Addend a) ASTM A1-68a Standard Specifications for Carbon Steel Rails ASTM A6-70 Specifications for General Requirements for Delivery of Rolled Steel Plates, Shapes, Sheet Piling and Bars for Structural Use ASTM A53-73 Standard Specifications for Welded and Seamless Steel Pipe ASTM A108-73 Standard Specifications for Cold Finished Carbon Steel Bars and Shafting ASTM A123-73 Standard Specification for Zinc (Hot Galvanized) Coatings on Products Fabricated from Rolled, Pressed and Forged Steel Shapes, Plates, Bars and Strip ASTM A143-74 Recommended Practice for Safeguarding Against Embrittlement of Hot-Dip Galvanized Structural Steel Products and Procedure for Detecting Embrittlement ASTM A153-73 Standard Specification for Zinc Coating (Hot

-Dip) on Iron and Steel Hardware ASTM A185-70 Specification for Welded Steel Wire Fabric for Concrete Reinforcement ASTM A193-75 Specification for Alloy Steel and Stainless Steel Bolting Material for High Temperature Service ASTM A194-74 Standard Specification for Carbon and Alloy Steel Nuts and Bolts for High Pressure and High Temperature Service ASTM A240-71 Specification for Stainless and Heat Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure Vessels

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 98 Code or Specification Title ASTM A276-75 Standard Specification for Stainless and Heat Resisting Bars and Shapes ASTM A307-68 Specification for Low Carbon Steel Externally and Internally Threaded Standard Fasteners ASTM A312-74 Standard Specification for Seamless and Welded Austenitic Stainless Steel Pipes ASTM A325-71 Specification for High Strength Bolts for Structural Steel Joints Including Suitable Nuts and Plain Temperature Service ASTM A358-75 Standard Specification for Electric

-Fusion-Welded Austenitic Chromium-Nickel Alloy Steel Pipe for High Temperature Service ASTM A370-75a Methods and Definitions for Mechanical Testing of Steel Products ASTM A384-72 Safeguarding Against Warpage and Distortion During Hot

-Dip Galvanizing of Steel Assemblies ASTM A385-62 Recommended Practice for Providing High Quality Coatings (Hot Dip) on Assembled Products ASTM A386-73 Standard Specification for Zinc Coatings (Hot Dip) on Assembled Products ASTM A391-65 Standard Specification for Alloy Steel Chains ASTM A394-75 Specification for Galvanized Steel Transmission Tower Bolts and Nuts ASTM A446-72 Standard Specification for Steel Sheet Zinc Coated (Galvanized) by the Hot

-Dip Process Physical (Structural) Quality ASTM A479-75 Standard Specification for Stainless and Heat

-Resisting Steel Bars and Shapes for Use in Boilers and Other Pressure Vessels ASTM A480-74 Standard Specification for Delivery of Flat

-Rolled Stainless and Heat Resisting Steel Plate, Sheet and Strip

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 99 Code or Specification Title ASTM A490-71 Specification for Quenched and Tempered Alloy Steel Bolts for Structural Steel Joints ASTM A501-74 Standard Specification for Hot

-Formed Welded and Seamless Carbon Steel Structural Tubing ASTM A502-75 Standard Specification for Structural Steel Rivets ASTM A514-7 0 Specification for High

-Yield Strength, Quenched and Tempered Alloy Steel Plate, Suitable for Welding ASTM A516-74a Specification for Pressure Vessel Plates, Carbon Steel, for Moderate and Lower Temperature Service ASTM A525-73 Standard Specification for General Requirements for Delivery of Steel Sheet, Zinc Control (Galvanized) by the Hot

-Dip Process ASTM A540-70 Alloy Steel Bolting Materials for Special Applications ASTM A570-72 Hot-Rolled Carbon Steel Sheet and Strip ASTM 572-75 High Strength Low Alloy, Columbium

-Vanadium Steels of Structural Quality ASTM A588-77a Specification for High

-Strength Low Alloy Structural Steel with 50,000 psi Minimum Yield Point to 4 in. Thick ASTM A615-72 Specification for Deformed Billet

-Steel Bars for Concrete Reinforcement ASTM A759-78 Standard Specification for Carbon Steel Crane Rails ASTM C29-71 Standard Methods of Test for Unit Weight of Aggregate ASTM C31-69 Standard Method of Making and Curing Concrete Compressive and Flexural Strength Test Specimens in the Field ASTM C33-71a Specification for Concrete Aggregates ASTM C39-71 Standard Method of Test for Compressive Strength of Molded Concrete Cylinders

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 100 Code or Specification Title ASTM C40-73 Standard Method of Test for Organic Impurities in Sands for Concre te ASTM C42-68 Standard Method of Obtaining and Testing Drilled Cores and Sawed Beams of Concrete ASTM C70-73 Standard Method of Test for Surface Moisture in Fine Aggregate ASTM C87-69 Standard Method of Test for Effect of Organic Impurities in Fine Aggregate on Strength of Mortar ASTM C88-73 Standard Method of Test for Soundness of Aggregates by Use of Sodium Sulfate or Magnesium Sulfate ASTM C94-72 Specification for Ready

-Mixed Concrete ASTM C109-73 Standard Method of Test for Compressive Strength of Hydraulic Cement Mortars (Using 2

-inch Cube Specimens)

ASTM C114-69 Standard Methods of Chemical Analysis of Hydraulic Cement ASTM C117-69 Standard Method of Test for Materials Finer than No. 200 (75 M) Sieve in Mineral Aggregates by Washing ASTM C123-69 Standard Method of Test for Light Weight Pieces of Aggregate ASTM C125-74 Standard Definitions of Terms Relating to Concrete and Concrete Aggregates ASTM C127-73 Standard Method of Test for Specific Gravity and Absorption of Coarse Aggregates ASTM C128-73 Standard Method of Test for Specific Gravity and Absorption of Fine Aggregate ASTM C131-69 Standard Method of Test for Resistance to Abrasions of Small Size Coarse Aggregate by Use of Los Angeles Machine ASTM C136-71 Standard Method of Test for Sieve or Screen Analysis of Fine and Coarse Aggregate

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 101 Code or Specification Title ASTM C138-75 Standard Method of Test for Unit Weight, Yield and Air Content (Gravimetric) of Concrete ASTM C142-71 Standard Method of Test for Clay Lumps and Friable Particles in Aggregates ASTM C143-71 Standard Method of Test for Slump of Portland Cement Concrete ASTM C150-71 Specification for Portland Cement ASTM C151-74a Standard Method of Test for Auto

-clave Expansion of Portland Cement ASTM C172-71 Standard Method of Sampling Fresh Concrete ASTM C173-75 Standard Method of Test for Air Content of Freshly Mixed Concrete by the Volumetric Method ASTM C186-73 Standard Method of Test for Heat of Hydration of Hydraulic Cement ASTM C191-74 Standard Method of Test for Time of Setting of Hydraulic Cement by Vicat Needle ASTM C192-69 Standard Method of Making and Curing Concrete Test Specimens in the Laboratory ASTM C231-71T Tentative Method of Test for Air Content of Freshly Mixed Concrete by the Pressure Method ASTM C233-73 Standard Method of Testing Air

-Entraining Admixtures for Concrete ASTM C235-68 Standard Method of Test for Scratch Hardness of Coarse Aggregate Particles ASTM C260-69 Specification for Air

-Entraining Admixtures for Concret e

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 102 Code or Specification Title ASTM C289-71 Standard Method of Test for Potential Reactivity of Aggregates (Chemical Method)

ASTM C294-69 Description of Nomenclature of Constituents of Natural Mineral Aggregates ASTM C295-65 Recommended Practice for Petrographic Examination o f Aggregates for Concrete ASTM C309-74 Standard Specification for Liquid Membrane

-Forming Compounds for Curing Concrete ASTM C404-70 Standard Specification for Aggregates for Masonry Grout ASTM C494-71 Specification for Chemical Admixtures for Concrete ASTM C496-71 Standard Method of Test for Splitting Tensile Strength of Molded Concrete Cylinders ASTM C535-69 Standard Method of Test for Resistance to Abrasion of Large Size Coarse Aggregate by Use of the Los Angeles Machine ASTM C566-67 Standard Method of Test for Total Moisture Content of Aggregate by Drying ASTM C666-75 Standard Method of Test for Resistance of Concrete to Rapid Freezing and Thawing ASTM D75-71 Standard Method of Sampling Aggregates ASTM D512-67 Standard Methods of Test for Chloride Ion in Industrial Water and Industrial Waste Water ASTM D1411-69 Standard Method of Test for Water

-Soluble Chloride Present as Admixtures in Graded Aggregate Road Mixes ASTM D1888-67 Standard Method of Test for Particulate and Dissolved Matter in Industrial Water ASTM E109-63 Standard Method for Dry Powder Magnetic Particle Inspection

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 103 Code or Specification Title ASTM E329-72 Recommended Practice for Inspection and Testing Agencies for Concrete as Used in Construction. (Articles 7, 8 and 9 do not apply.) AISC Specification for the Design, Fabrication and Erection of Structural Steel for Building 1969 Edition (including Supplements 1, 2 and 3)

ANS 20.1 (ANSI N177)

Proposed Standard for the Design Basis for the Protection Against Internal and External Plant Missiles (April 1974)

ANSI A58.1

-1972 American Standard Building Code Requirements for Minimum Design Loads in Buildings and Other Structures ANSI B31.1

-1973 Summer and Winter Addenda, Power Piping ANSI N45.2

-1974 Quality Assurance Program Requirements for Nuclear Power Plants ANSI N101.2

-1972 Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities ANSI N101.4

-1972 Quality Assurance for Protective Coatings Applied to Nuclear Facilities ANSI N101.6

-1972

  • American National Standard for Concrete Radiation Shield ANSI N512-1974 Protective Coating (Paints) for Nuclear Industry AWS B3.0-41 Standard Qualification Procedures AWS D1.0-69 Standard Code for Arc and Gas Welding in Building Construction

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 104 Code or Specification Title AWS D1.1-75 Structural Welding Code AWS D1.1-75 Recommended Practice for Welding Reinforcing Steel, Metal Inserts and Connections in Reinforced Concrete Construction American Railway Engineering Association (AREA) Manual of Railway Engineering, Volumes I & II (1972 Revision)

American Hot Dip Galvanizing, Inc.

Hot Dip Galvanized Coatings (1973)

ASNT SNT-TC-1A (June, 1975) American Society for Non

-Destructive Testing Recommended Practice for Non

-Destructive Testing

-Personnel Qualifications and Certification Bethlehem Steel Crane Rails Catalog No. 3351 (updated to December 1979)

Bethlehem Steel Specification for Fully Heat Treated Rails (March 1977 Revision)

Mixer Manufacturers Bureau of the Associated General Contractors of America Concrete Plant Mixer Standards (1974 Issue)

MMB of the AGCA Recommended Guide Specification for Batching Equipment and Control Systems in Concrete Batch Plants Publication 102 (1974 Issue) NRMCA National Ready

-Mixed Concrete Association Certification of Ready-Mixed Concrete Production Facilities Instructions and Check List (1972)

NRMCA Truck Mixer and Agitator Standards of Truck Mixers Manufacturer's Bureau (1971)

U.S. Army Corps of Engineers Specification

CRD-C119-63 Method of Test for Flat and Elongated Particles in Coarse Aggregate SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 105 Code or Specification Title U.S. Dept. of Labor OSHA Occupational Safety and Health Administration Standards, 1975 Edition U.S. Dept. of Commerce National Bureau of Standards PS-1-74 Construction and Industrial Plywood U.S. Dept. of Commerc e National Bureau of Standards Handbook 44, Specifications, Tolerances and Other Technical Requirements for Commercial Weighing and Measuring Devices, 1971 SSPC Steel Structures Painting Council Steel Structures Painting Manual, Volume 2; Systems and Specifications including supplement, 1973 SSPC Good Painting Practice, Volume 1, 1966 Uniform Building Code International Conference of Building Officials, Uniform Building Code, 1973 Edition NRC TID 7024 Nuclear Reactors and Earthquakes, (August 1 973) NRC 10 CFR 50 App. B Quality Assurance Criteria for Nuclear Power Plants Nuclear Construction Issues Group (NCIG

-01, Rev. 2, 5/7/85)

Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 106 NRC Regulatory Guide No.

Title 1.10, Rev. 1, 1/73 Mechanical (Cadweld) Splices in Reinforcing Bars of Concrete Containments 1.15, Rev. 1, 12/72 Testing of Reinforcing Bars for Concrete Structures 1.29, Rev. 3, 9/78 Seismic Design Classification 1.54, Rev. 0, 8/72 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants 1.55, Rev. 0, 6/73 Concrete Placement in Category I Structures 1.60, Rev. 1, 12/73 Design Response Spectra for Seismic Design of Nuclear Power Plants 1.61, Rev. 0, 10/73 Damping Values for Seismic Design of Nuclear Power Plants 1.69, Rev. 0, 12/73 Concrete Radiation Shields for Nuclear Power Plants 1.82, Rev. 0, 6/74 Sumps for Emergency Core Cooling and Containment Spray Systems Revision 3 was used for guidance for the replacement sump strainers along with NEI 04

-07. 1.92, Rev. 1, 2/76 Combining Modal Responses and Spatial Components in Seismic Response Analysis 1.94, Rev. 1, 4/76 Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During Construction Phase of Nuclear Power Plants 1.142, Rev. 0, 4/78 Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments)

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 107 The below listed UE&C design and construction specifications applicable to the containment internal structures and to the other seismic Category I structures were prepared in accordance with applicable codes, quality control requirements, and NRC Regulatory Guides:

UE&C Specification Title 9763.006-12-1 Structural Steel Work 9763.006-12-2 Structural Steel Erection 9763.006-12-5 Fabrication of Safety

-Related Structural Steel Work 9763.006-13-3 Category I Concrete Work Other Than Containment 9763.006-14-1 Furnishing, Detailing, Fabricating, and Delivering Reinforcing Bars 9763.006-14-3 Installation of Reinforcing Bars in Category I Structures (Other Than Containment) 9763.006-18-1 Furnishing of Miscellaneous Embedded Steel and Weldments 9763.006-18-2 Installation of Miscellaneous Steel and Weldments 9763.006-18-3 Furnishing of Miscellaneous Steel 9763.006-18-4 Furnishing and Installing Embedded Steel and Miscellaneous Steel 9763.006-41-4 Furnishing of Protective Coating (Paint) System Materials and Related Services 9763.006-41-7 Field Painting of Containment Structure Interior 9763.006-69-1 Concrete Batch Plant 9763.006-69-3 Concrete Mixes 9763.006-69-6 Ready-Mixed Concrete from Off

-Site Plant 9763.006-69-7 Standard Concrete Mixes 976 3.006-69-10 Ready-Mixed Concrete for Category I, Non-Category I Structures and Systems from an Off

-Site Plant

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 108 UE&C Specification Title 9763.006-225-2 Stainless Steel Liners 9763-MPS-1 Material and Processing Requirements (Nuclear) 9763-MPS-2 Material and Processing Requirements (Non

-Nuclear) 9763-MPS-3 Material and Processing Requirements of Welded Studs, Reinforcing Bars and Anchor Bolts 9763-QAS-1 Administrative and System Requirements (Nuclear) 9763-QAS-2 Administrative and System Requirements (Non

-Nuclear) 9763-RM-1 Instructions for Site Records Management System 9763-WS-2 Requirements for Welding and Non

-Destructive Examination for Non-Nuclear Pressure Components and Non

-Nuclear Power Piping 9763-WS-3 Requirements for Welding and Non

-Destructive Examination for Structural Steel 9763-WS-4C Requirements for Cadwelding and Non

-Destructive Examination of Mechanical Rebar Splice Method 3.8.3.3 Loads and Loading Combinations The containment internal structures are designed to withstand all credible conditions of loading including normal loads, severe environmental loads, extreme environmental loads, and abnormal plant loads. These loads are determined in accordance with the applicable specifications, including ACI 318

-71 and AISC

-69, and are considered in Normal and Unusual Load Combinations to assure that the response of the structure remains within the limits prescribed in Subsection 3.8.3.5. Refer to Subsection 1.8 (Regulatory Guide 1.142) for a discussion concerning compliance with ACI

-349.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 109 a. Design Loads The definitions of the loads used in the design of the internal structures include the following:

1. Normal Plant Startup, Operation and Shutdown Loads Normal loads are those loads encountered during normal plant start

-up, operation and shutdown.

They include the following:

(a) Dead Loads (D) Dead loads are all permanent gravity loads including the weight of concrete walls and slabs, structural framing, piping, cable and cable trays, permanent equipment, and static pressures of liquids.

(b) Live Loads (L) Live loads include any movable equipment loads and other loads which vary in intensity and/or occurrence. Live loads are present only during shutdown conditions, and do not govern the design of any components.

(c) Operational Thermal Loads (T o) The temperature gradient through the walls under normal operating conditions is considered in the design. For a discussion of minimum and maximum operating temperatures, see Subsection 6.2.1. (d) Operational Pipe Reactions (R o) These are pipe reactions due to thermal conditions existing in the piping during normal operation or shutdown. They are based on the most critical transient or steady

-state condition.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 110 2. Severe Environmental Loads Severe environmental loads are those loads that result from events that could infrequently be encountered during the plant life. The only load included in this category is the following:

(a) Operating Basis Earthquake (E o) These are the loads generated by the Operating Basis Earthquake, which is the earthquake that could reasonably be expected to affect the site during the operating life of the plant. Only the actual dead load and weights of fixed equipment are considered in evaluating the seismic response forces.

The horizontal and vertical design response spectra for the OBE are derived by applying a factor of 0.5 to the response spectra given for the Safe Shutdown Earthquake (SSE) which is described below. The effects of two (2) orthogonal horizontal components and one (1) vertical component of earthquake are considered and combined by the square

-root-of-the-sum-of-the-squares rule.

3. Extreme Environmental Loads Extreme environmental loads are those loads which result from events which are credible but highly improbable. The only load included in this category is the following:

(a) Safe Shutdown Earthquake (E ss) These are the loads generated by the Safe Shutdown Earthquake, which is the earthquake based upon an evaluation of the maximum earthquake potential in the vicinity of the plant. Dead and fixed equipment loads are described under the Operating Basis Earthquake, above. The horizontal and vertical forces on the internals are developed from the response spectra given in Figure 2.5-38 and Figure 2.5-39 the development of which is described in Subsection 2.5.2.6. The effects of two (2) orthogonal horizontal earthquakes and one (1) vertical earthquake are considered and combined by the square-root-of-the-sum-of-the-squares rule.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 111 4. Abnormal Loads Abnormal loads are those loads generated by a postulated high energy pipe break resulting in a loss

-of-coolant accident (LOCA). The design basis accident results in the highest postulated pressures and temperatures, and is determined by considering a LOCA.

(a) Accident Pressure (P a) Immediately following a postulated primary pipe break and prior to pressure equalization, a differential pressure occurs on the interior structures. See Subsection 6.2.1.2 for a description of these pressure loads, including the effects of the neutron shielding on pressures. Note that the differential pressure load on the primary shield wall is greater than the differential pressure load for the secondary shield wall; however, in this subsection one symbol is used to represent this loading with the understanding that the value is as specified in Subsection 6.2.1.2. Accident pressures resulting from a main steam line break are also considered.

(b) Accident Temperature (T a) Also following a postulated primary pipe break is an increase in temperature on the interior structures. These loads are also further described in Subsection 6.2.1.3. (c) Accident Pipe Reactions (R a) Pipe reaction loads due to thermal conditions generated by the postulated pipe break, including R o, are considered in the design. The magnitude of these loads is determined by the piping design.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 112 (d) Pipe Break Loads (R r) Pipes, other than the Primary Coolant System, are anchored in sleeves through the interior concrete structures and transmit thrust loads to the structure during normal operating conditions and for the postulated pipe break. In addition, loads transmitted by pipe whip restraints are considered. The three components of pipe break loads are as follows:

(1) R rr = load on the structure generated by the reaction of a ruptured high energy pipe during the postulated pipe break. The time-dependent nature of the load and the ability of the containment to deform beyond yield are considered in establishing the structural capacity necessary to resist the effects of R rr. (2) R rj = load on the structure generated by jet impingement from a ruptured high energy pipe during the postulated pipe break. In general, direct impingement of steam on the structure does not produce significant design loadings due to the distance between the wall and the break location.

Where a break is postulated to occur close enough to a wall to produce a critical loading, a shield, or deflector, is provided, and the loading is transferred to the embedment for the pipe whip restraint to which the shield is attached.

(3) R rm = load on the containment resulting from the impact of a ruptured high energy pipe during the postulated pipe break. Since all high energy lines are constrained by pipe rupture restraints, missile loadings of this nature are prevented.

5. Other Loads and Load Considerations (a) Internal Missile Loads (M) Internal missile loads, other than those defined as R rr , R rj, and R rm , are considered including an appropriate dynamic factor to account for the dynamic nature of the load. For a discussion of specific missiles generated inside the containment and the missile design procedures, see Section 3.5. Note that the sides and floor of the refueling canal are designed for the impact of the neutron shielding.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 113 (b) Primary Equipment Supports The primary equipment supports transmit loads to the fill mat, to the primary and secondary shield walls and to the operating floor slab. For a discussion of these loads and of the design criteria for the supports see Subsection 5.4.14. (c) Load Considerations for Internal Structures (1) Time dependent loads such as thermal effects, creep and shrinkage do not have any significant effects on the internal structures since the accident loads are generally resisted in tension by the reinforcing steel. In addition, the accident loads are short

-term, once-occurring loads, which have negligible creep effects.

(2) The containment structure functions as an independent structure with complete physical separation from the internal structures, and therefore there are no loading interactions between the two.

(3) Compartmentalization is considered in the design of the internal structures by using the peak subcompartment differential pressures, plus a safety margin. This is further discussed in Subsection 6.2.1.2. b. Load Combinations Various load combinations are considered in design to determine the greatest strength requirements of the structure. Where varying loads occur, the combinations producing the most critical loading are used. Basic combinations in the design of the containment internal structures are given in Table 3.8-14. These load combinations are in agreement with Subsections II.3 and II.5 of the Standard Review Plan for Subsection 3.8.3 of the UFSAR. The factors which are to be applied to allowable stresses have been transposed and applied as load factors instead, resulting in acceptance criteria as indicated in the table. Two categories of loading conditions and criteria are used in the design of the containment internal structures as described below.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 114 1. Normal Load Conditions Normal load conditions e those encountered during testing and normal operation. They include dead load, live load and anticipated transients or test conditions during normal and emergency startup and shutdown of the Nuclear Steam Supply, Safety and Auxiliary Systems. Normal loading also includes the effect of an Operating Basis Earthquake. Normal load conditions are referred to in Division 2 as service load conditions.

Under each of these loading conditions, the structure is designed so that stresses will be within the elastic limits. Design assumptions are presented in Subsection 3.8.3.4 and stress limitations are presented in Subsection 3.8.3.5. 2. Unusual Load Conditions Unusual load conditions are those conditions resulting from combinations of the LOCA, SSE and OBE, high-energy pipe failures, and live and dead loads. They are referred to in Division 2 as factored load conditions.

For each of the unusual loading combinations, the internal structures are designed to remain below their ultimate capacity so that the behavior of structural components is in the small deformation elastic range. Design assumptions are presented in Subsection 3.8.3.4 and stress limitations are presented in Subsection 3.8.3.5. 3.8.3.4 Design and Analysis Procedures

a. Design and Analysis After preliminary design, the internal structures are analyzed to determine the maximum stress and displacements in reinforcing steel, concrete and structural steel for the loading criteria described in Subsection 3.8.3.3. Each structural system, either an individual component or a group of connected components, is analyzed separately using appropriate design assumptions and boundary conditions.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 115 The analyses include fixed base conditions, uncracked concrete, load superposition, modeling and selective interpretation of results. Material properties and boundary conditions are selected to be representative of the particular structure or component being analyzed. Either conventional methods and formulae of structural mechanics or the finite element direct stiffness method are used. All analyses consider elastic behavior only. The design is based on codes, specifications and documents listed in Subsection 3.8.3.2. The methods of design and analysis, including assumptions, boundary conditions, loads resisted and analysis techniques are given below. Seismic analysis is discussed in Subsection

3.7.2. Reinforcing

in concrete components is proportioned in accordance with the responses obtained from the analyses using the strength design method of the ACI 318-71 code, as applicable, and the requirements of Subsections 3.8.3.3 and 3.8.3.5. Bond and anchorage requirements are in accordance with ACI 318-71. As a result, reinforcing patterns in walls and floors, in general, consist of orthogonal layers on both the inside and outside faces of the concrete.

The computer program CONCOL (described in Subsection 3.8.3.4) was used throughout the internals for ultimate strength design of concrete. Other computer programs used in the design and analysis of specific internal structures are described below.

Structural steel is designed in accordance with the AISC Specification, using elastic methods, except as noted below for the reactor support system.

Refer to Subsection 1.8 (Regulatory Guide 1.142) for a discussion concerning compliance with ACI

-349. b. Internal Structures

1. Reactor Support System Design and analysis of the vertical supports are described in Subsection 5.4.14. The reactor vessel lateral support, the ring girder, is analyzed using a three-dimensional finite element model. Since the support possesses bilateral symmetry and is subjected only to unilateral horizontal loading, only half of the support was required for the model. Loadings used in the analyses were supplied by the NSSS vendor, Westinghouse.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 116 Design and analysis procedures for the reactor vessel lateral support are in accordance with the ASME Boiler & Pressure Vessel Code, Division 1, Subsection NF and Appendix XVII. 2. Steam Generator Support System Design and analysis of the steel supports are described in Subsection 5.4.14. 3. Reactor Coolant Pump Supports Design and analysis of the reactor coolant pump supports are described in Subsection 5.4.14. 4. Primary Shield Wall The reactor pressure vessel is enclosed and supported by a reinforce d concrete primary shield wall. The primary shield wall is designed under the accident condition considering the effects of temperature, jet forces and forces transmitted by the reactor vessel, seismic loads, etc. The design pressure loads are conservatively considered as peak transient pressure differential on the wall. Other LOCA forces on the structure are determined using dynamic time

-dependent analysis. The peak values of these loads are conservatively used as equivalent static loads.

The wall is analyzed as an axisymmetric structure resting on the fill mat using the WILSON 2 computer code (described in Subsection 3.8.1.4). The analysis is performed assuming base fixity and top restraint provided by the floor of the refueling canal. Differential pressure loads and thermal loads are analyzed as axisymmetric load cases. Differential pressures are considered as nonaxisymmetric loads and are represented by a Fourier series. An additional consideration is the effect of radiation in generating heat in

the primary shield wall. This wall is the only concrete subjected to relatively high irradiation. The attenuation of the integrated neutron and gamma rays within the wall will produce a temperature rise which will be limited to a maximum of 150F at the concrete surface adjacent to the reactor vessel by the convection of cooling air circulated between the wall and the reactor vessel.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 117 Studies on the behavior of concrete under irradiation (Reference 5) indicate that irradiated specimens behave in a manner similar to control specimens subjected to temperature without radiation. Thus, the effects on the shield wall concrete from heat generated by irradiation were evaluated in the same manner as temperature rise occurring in the concrete from a nonradiating source. 5. Secondary Shield Wall The secondary shield wall is analyzed as part of a larger three

-dimensional model which includes the secondary shield wall, the refueling canal, the operating floor slab, the crane support structure and the biological shield wall. The computer program used was STARDYNE (described in Subsection 3.8.1.4). The model is composed of plate elements (which include bending and transverse shear) for the walls and slabs and beams for the crane support structure and other uses as required. Restraint is provided at the level of the fill mat and at the interfaces between the operating floor and the primary shield wall. Differential pressures, thermal loads, pipe break loads, and normal operating pipe loads are applied to the finite element model. Force and displacement boundary conditions imposed by the primary shield wall onto the operating floor slab are considered.

6. Refueling Canal The refueling canal is analyzed as part of the model described above under the heading, "Secondary Shield Wall." 7. Pressurizer Region The pressurizer region is analyzed for the established load criteria using an independent three dimensional finite element model and the general procedures described above under the heading, "Secondary Shield Wall." 8. Control Rod Drive Mechanism (CRDM) Missile Shield The CRDM missile shield has been integrated into the simplified head assembly. The missile shield has been evaluated to the criteria presented in the Standard Review Plan (SRP) Section 3.5.3. The missile shield and supporting structure have been modeled using the ANSYS computer code.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 118 9. Operating Floor Slab The operating floor slab is analyzed as part of the model described above under the heading, "Secondary Shield Wall." 10. Fill Mat The fill mat is analyzed as an independent circular reinforced concrete slab; the model consists of plate elements (which include bending and transverse shear). It is designed to resist and transfer to the containment base mat all reaction loads from the internal structures including the primary equipment supports. The internal structures are generally rigidly connected to the fill mat.

11. Structural Steel Structural steel framing and platforms, including connections, are conventionally designed using applicable codes and specifications, as listed in Subsection 3.8.3.2. Those members which support piping, electrical cable trays and ducts were further evaluated for rigidity considerations as described below.

All members of the containment annulus steel framing that support piping or piping and electrical cable trays and/or ducts were initially designed to have natural frequencies of 20 hz or greater, in order to minimize the seismic loads in the supported components. For members supporting only ducts and/or electrical cable trays the minimum frequency was 15 hz. Where meeting these criteria is not practicable, members supporting piping were initially designed for a minimum frequency of 15 hz, if practicable.

Members that were not sized using the frequency approach were initially designed for static loads and an additional 2.0 g factor (horizontal and vertical).

Final member sizes were all checked for conformance with applicable codes against final loadings obtained from dynamic analyses of steel framing, piping, ducts and equipment.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 119 12. Crane Support Structure The crane support structure is analyzed as part of the model described above under the heading, "Secondary Shield Wall." 13. Control Rod Drive Mechanism (CRDM) and Drive Rod Position Indication (DRPI) Cable Bridges To support the CRDM and DRPI cables that run to the simplified head assembly, two hinged cable bridges have been mounted on the operating deck, one on each side of the refueling cavity. The bridges have been modeled using the STAAD III computer code. c. Description of Computer Programs Utilized in the Design and Analysis The computer programs used in the design and analysis of the containment internal structures are described briefly in this subsection. Programs listed below are only those programs which have not been described previously in Subsections 3.8.1.4g and 3.8.2.4i. A summary of the comparisons of results used to validate them is given in Appendix 3F. The program classifications are discussed in Subsection 3.8.1.4g. 1. CONCOL (SYSTEM PROFESSIONAL): Documentation is available from Control Data Corporation. CONCOL is a computer program for ultimate strength design of concrete columns.

2. STRAP: "Static Analysis of Linear Elastic Structures," by United Engineers & Constructors, Inc. STRAP is used to perform static analyses of structures which can be represented as an assemblage of linear elastic members. It uses the stiffeners matrix methods of analysis.
3. GENSAP (STRU

-PAK): "Static Analysis of Elastic Structures." Documentation is available from Control Data Corporation. GENSAP is used for general static analysis of elastic structures composed of beams and columns.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 120 4. STAAD III Structural Analysis and Design Software:

A proprietary computer program of Research Engineers, Inc. (REI), California, for the analysis and design of structures. The code has been placed under configuration control by Westinghouse and specific features of this software utilized in the simplified head assembly calculations have been independently verified by Westinghouse. (Westinghouse Letter Number EDRE-CSE-134(97), Software Release of STAAD III (22.0W) on Windows NT System, 9/25/97, and Westinghouse Calculation

  1. CSE-06-98-0001, Rev. 0, titled: "STAAD Verification Problem-Response Spectra Analysis.")
5. GOTHIC Generation of Thermal

-Hydraulic Information for Containments:

GOTHIC is a general purpose thermal

-hydraulic computer program for design, licensing, safety and operating analysis of nuclear power plant containments and other confinement buildings. GOTHIC has been developed for the Electric Power Research Institute (EPRI) by Numerical Applications Inc, Richland, Washington. The code has been placed under configuration control by Westinghouse. (GOTHIC Qualification Report for Version 5.0e (NAI 8907

-09, Rev. 3, Dec. 1995.)

3.8.3.5 Structural Acceptance Criteria The bases for the development of the following stress

-strain criteria are the ACI 318-71 and AISC codes.

a. Normal Load Conditions Internal structures are proportioned to remain within the elastic limits under all normal loading conditions described in Subsection 3.8.3.3. Reinforced Concrete

- designed in accordance with ACI 318-71 Strength Method, which insures flexural ductility by control of reinforcing steel percentages and stresses. Structural and Miscellaneous Steel

- designed in accordance with AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings, Part I.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 121 b. Unusual Load Conditions Internal structures are designed to maintain elastic behavior under all unusual load conditions shown in Subsection 3.8.3.3. The upper bound of elastic behavior is taken as the yield strength capacity of the load carrying components. The yield strength of structural and reinforcing steel is taken as the minimum guaranteed

yield stress as given in the appropriate ASTM Specification.

Reinforced Concrete

- designed in accordance with ACI 318-71 Building Code. Member yield strength is considered to be the strength capacity calculated by the ACI Code. Structural and Miscellaneous Steel - designed in accordance with AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings, Part

1. Overall stability of steel structures designed for unusual loading is verified using the AISC Specification, Par t 2, and the load factors in Table 3.8-14, with the exception of the reactor vessel lateral support, which is designed in accordance with ASME B&PV Code, Division 1, Subsection NF. c. Deformations Each of the structures is designed to remain in the small deformation elastic range so that no gross deformations will occur and cause contact with other structures or pieces of equipment.
d. Seismic Analysis The seismic analysis techniques used for seismic Category I structures and systems are described in Subsection 3.7.2. The seismic analysis techniques used for seismic Category I subsystems are described in Subsection 3.7.3. The shear produced by the seismic response of the internal structures is accounted for in design. If the available shear strength of the concrete section is not adequate, shear reinforcement is used to increase the shear resistance of the section. In general, seismic loads are not the governing factor in the design of the internal structures. Consequently, shear produced by seismic response is resisted by orthogonal layers of reinforcing provided for the applicable design loading combination; diagonal seismic reinforcing is not required. The shear response of the internal structures is transferred to the fill mat which is keyed into the containment base mat at the reactor cavity pit.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 122 3.8.3.6 Materials, Quality Control and Special Construction Techniques The primary materials of construction for the internal structures are concrete, reinforcing steel, structural steel (rolled shapes and plates), stainless steel (liner plate), and field coatings. Descriptions of these materials and respective quality control are discussed below. There are no special construction techniques.

a. Concrete Concrete work is in accordance with ACI 318-71 and ACI 301 codes, except as noted in Subsection 1.8 (Regulatory Guide 1.55). The concrete is a dense, durable mixture of sound, coarse aggregate, cement and water. Admixtures were added, where required, to improve the quality and workability during placement and to retard the set of the concrete. Engineering approval was required prior to the use of admixtures.

Aggregate conforms to ASTM C33. It consists of inert materials that are clean, hard and durable, free from organic material and uncoated with clay or dirt. Fine aggregate consists of natural sand and the coarse aggregate consists of crushed stone. Portland cement conforms to ASTM C150, Type II (moderate heat of hydration requirements).

Water is clean and free from any deleterious amounts of acid, alkali, salts, oil, sediment, organic matter or other substances which may be harmful to the concrete or steel.

The reinforced concrete has a nominal density of 150 lb/ft 3, which is used for determination of dead load. Shielding calculations for the primary shield wall are based on a dry concrete density of 139 lb/ft 3; other shielding calculations are based on a dry concrete density of 147 lb/ft 3. The 28-day standard compressive strength of the concrete is 4000 psi. To assure that adequate means of control were used in the manufacture and that the properties described above were realized, the following were required:

1. Suppliers, fabricators and contractors were required to have written quality assurance procedures, which were reviewed and approved by United Engineers. Material certifications were required in accordance with the applicable portions of the quality assurance plan described in Chapter 17 of the UFSAR and in the material specifications.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 123 2. The following tests and certifications were required: (a) Aggregates were tested to comply with ASTM C33. (b) Cement was tested in accordance with ASTM C114 to conform to ASTM C150. (c) Concrete samples were taken from the mix in accordance with ASTM C172. (d) Cylinders were made in accordance with ASTM C31. (e) Compressive tests were made in accordance with ASTM C39. (f) Slump tests were in accordance with ASTM C143 and air content tests were in accordance with ASTM C231. (g) Evaluation of the tests was in accordance with the material specifications, the codes discussed above, and the ACI 318

-71 code, as applicable.

3. All making and testing of concrete samples was accomplished by an independent testing laboratory.

All concrete operations during cold weather conditions followed the practice defined in ACI 301 and 306R

-78. During cold weather curing of the concrete, concrete surfaces whose temperatures are below 50F by accident for short periods of time, but remain 40F or above, have had the 7

-day curing period extended by the amount of time the concrete was below 50F (rounded out to the nearest whole day).

The strength maturity method as defined in ACI 306R-78 was allowed as an alternate to specified curing requirements. The required maturity factor was selected to insure that concrete attained 7 0 percent of the design strength. This method was also used to determine the attainment of

in-place concrete strength required for removal of forms in lieu of field cured cylinder as required in ACI 301.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 124 b. Reinforcing Steel Reinforcing steel consists of high-strength deformed billet bars conforming to ASTM A615, Grade

60. Certified Material Test Reports and user tests, as required by Regulatory Guide 1.15, were provided by the material manufacturer.

All reinforcing bars were spliced in accordance with UE&C Specification 9763-WS-4C, and No.

14 and 18 bars were joined by mechanical butt splices (Cadweld splices).

Additional information on requirements for reinforcing steel can be found in Subsection 3.8.1.6. c. Structural Steel All structural and miscellaneous steel, including stainless, conforms to the following specifications unless otherwise noted.

Structural steel conforming to ASTM A36 was used except where high strength is required. ASTM A588 or A572, Grade 50 was used for these applications, suc h as the reactor vessel lateral support, the cross

-over leg restraint and the hot leg restraint.

To assure that steel plates, such as embedded plates for pipe restraints, are able to transmit orthogonal loads, a material with good through

-plate tensile properties, ASTM A36 steel, was used.

Bolts were made of ASTM A325 steel except in special applications. Where high strength was required ASTM A193, A490, or A540 bolts were used. ASTM A307 bolts were used for stair stringers, stair treads, grating, handrail, toe plates, girts and purlins.

Stainless steel liner plate was fabricated from material conforming to ASTM A240, TP-304 steel. A liquid penetrant test is performed on the liner plate on the refueling canal to assure leak tightness.

The crane rail is a 171 lb/ft rail fabricated from A

-759 steel; it was fully heat treated. All welding conforms to either the American Welding Society Code (AWS) or to the ASME Boiler and Pressure Vessel Code (Section III).

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 125 For structural steel governed by AWS D1.1 and not subject to fatigue loads, welds shall be visually inspected in accordance with NCIG

-01, "Visual Weld Acceptance Criteria (VWAC)," as specified in UE&C Specification WS

-3. The only exception is for structural steel governed by AWS D1.1 which are subject to fatigue; no undercut is permitted.

Welders of carbon steel are qualified in accordance with the "Standard Qualification Procedure" of the AWS; welders of stainless steel liner plate are qualified in accordance with Section IX of the ASME Boiler and Pressure Vessel Code. Certified Material Test Reports giving chemical composition and physical properties were supplied by the manufacturers for all structural steel.

Fabrication and erection was in accordance with AISC standards and the applicable material specifications.

d. Steel and Concrete Coating System Materials used for coating the internal structures are the same as those described in Subsection 3.8.1.6. These materials meet the requirements of ANSI Standard N101.2. 3.8.3.7 Testing and In

-Service Surveillance Requirements Quality control testing as discussed in Subsection 3.8.3.6 will be employed. No additional testing or in-service surveillance is required.

3.8.4 Other

Seismic Category I Structures This section contains physical descriptions, codes, loads and load combinations, design and analysis procedures, allowable stresses, quality control, and testing requirements as they relate to seismic Category I structures exclusive of the containment structure and its internals. Nonseismic Category I structures which received special design considerations to prevent collapse onto adjacent Category I structures are also included in this section.

3.8.4.1 Description of the Structures Locations of all of the major plant structures are shown on the plant arrangement plan, Figure 1.2-1; overall dimensions of the seismic Category I structures are given in Table 3.8-15.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 126 All seismic Category I structures are separated from adjacent structures above the point of fixity by means of isolation joints, except for a few structures described below in which two portions of the same structure are used for two different functions. The resulting multi

-function structure is isolated from adjacent structures. See Subsection 3.8.5.1 for a detailed discussion of seismic isolation. The seismic Category I structures are described in the following paragraphs. They incorporate no unique or new design or construction features. Also, block or concrete masonry partitions are not utilized in any Category I structure.

a. Containment Enclosure Building The Enclosure Building is a reinforced concrete right cylindrical structure with a hemispherical dome. The inside diameter of the cylinder is 158 feet. The vertical wall varies in thickness from 36 inches to 15 inches; the dome is 15 inches thick. The inside of the dome is 5'

-6" above the top of the containment dome.

Located on the outside of the Enclosure Building is the plant vent stack, consisting of a light steel frame with steel plates and varying in cross section. The stack carries exhaust air from various buildings.

The space between the containment and the Enclosure Building is maintained at a slight negative pressure during accident conditions. All joints and penetrations are caulked or gasketed to ensure air tightness. See Figure 1.2-2, Figure 1.2-3, Figure 1.2-4, Figure 1.2-5 and Figure 1.2-6 for sections and elevations of the Enclosure Building.

b. Containment Equipment Hatch Missile Shield This shield is a removable, precast, reinforced concrete wall located outside th e equipment hatch. It serves to protect the hatch from tornado

-generated missiles. See Figure 1.2-4 and Figure 1.2-5 for a plan and a section.

c. Containment Enclosure Ventilation Area The containment enclosure ventilation area is an irregularly shaped reinforced concrete building that houses ventilation equipment (fans, filters, etc.) for the Enclosure Building and is located on the southwest side of the containment. Its overall dimensions are 116.0 feet long, 55.75 feet wide at its widest point, and

3 1.5 feet high. See Figure 1.2-2, Figure 1.2-3, and Figure 1.2-4.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 127 d. Control and Diesel Generator Building The Control and Diesel Generator Building is a reinforced concrete structure approximately 233 feet long by 90 feet wide. All load carrying walls a nd columns are founded on fill concrete and rock below grade. This is a multi-function structure in which the two portions, the control room area and the diesel generator area, are separated by a common wall in the north

-south direction and are not seismically isolated. The building was analyzed and designed as a unit. The east portion of the structure (Control Building), which is 138 feet long, has three floors and extends from grade to approximately 79 feet above grade. The two intermediate floors and roof are supported on steel columns in the center and on concrete walls all around. The ground floor contains electrical switchgear, motor generator sets and battery rooms; the second floor is for cable spreading; and the third floor is the main control room. The west portion of the structure (Diesel Generator Building), which is 95 feet long, has two floors and extends from 36 feet below grade to approximately

59 feet above grade. The portion below grade houses storage tanks for diesel fuel.

The area between elevations 20'

-0" and 50'

-0" is divided, north and south, by a 2 feet thick reinforced concrete wall which supports the second floor and provides protection for each diesel generator against missiles generated by the other. The second floor contains air intakes for the diesel generators and building ventilation equipment. The roof is supported by concrete walls all around and by steel columns in the center extending from the second floor and located directly over the dividing missile wall below.

See Figure 1.2-31, Figure 1.2-32, and Figure 1.2-33, Figure 1.2-34, Figure 1.2-35 and Figure 1.2-36 for plans and sections of the Control Building and Diesel Generator Building, respectively.

e. Control Room Makeup Air Intake Structure The control room makeup air intake structures are reinforced concrete structures which serve as terminals for buried ductwork that provides air for the control rooms during accident conditions. The intake is located near the demineralized water tank, north of the Circulating Water Pumphouse.

The top of the sleeve for the air intake is located at Elevation 22'-7", higher than the maximum flood elevation.

See Figure 3.8-31 for a plan and section.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 128 f. Emergency Feedwater Pump Building Including Electrical Cable Tunnels and Penetration Areas (Control Building to Containment)

The Emergency Feedwater Pump Building is a reinforced concrete building on the north end of the Enclosure Building. It extends to approximately 28 feet north of the Enclosure Building and is 84 feet wide.

In elevation, the building extends from Elevation (-)26'-0" (top of floor slab) to elevation 47'-0" (top of roof slab). The building consists of the emergency feedwater pump room located above a two-story high electrical cable tray tunnel. The roof is supported on three exterior walls and on columns, on the south side. Portions of the emergency feedwater pump rooms which extend beyond the electrical cable tray tunnels are supported on walls founded on rock or on concrete fill.

The emergency feedwater pump room contains emergency feedwater pumps, demineralized water makeup pumps, valve stations and an auxiliary control panel.

A monorail is provided for servicing the pumps. The electrical penetration areas are approximately 84 feet wide and are situated one on top of the other. They penetrate the Enclosure Building and join with the containment structure.

See Figure 1.2-2, Figure 1.2-3, Figure 1.2-4, and Figure 1.2-6, and Figure 1.2-51. g. Enclosure for Condensate Storage Tank The enclosure is a cylindrical reinforced concrete wall, 2 feet thick and 43 feet high, which surrounds each tank and extends up to the springline of the tank, providing protection from horizontal tornado

-generated missiles. The enclosure is capable of retaining the contents of the tank should the tank rupture due to a missile penetrating it from above. See Figure 3.8-32. h. Fuel Storage Building The Fuel Storage Building is a reinforced concrete structure approximately

98 feet square that extends approximately 44 feet below grade to 66 feet above grade. The building contains the new fuel storage area and the spent fuel pool.

The spent fuel pool is constructed of concrete walls of a minimum thickness of

6 feet and a 4'

-10" thick concrete floor; for leak tightness the inside surface is lined with stainless steel plates 3/16" thick on the walls and 1/4" thick on the floor. The building superstructure is constructed of concrete walls and roof.

New and spent fuel storage and handling are described in Section

9.1. Refer

to Figur e 1.2-15, Figure 1.2-16, Figure 1.2-17, Figure 1.2-18, Figure 1.2-19, Figure 1.2-20 and Figure 1.2-21 for plan and sectional elevations of the building.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 129 i. Main Steam and Feedwater Pipe Chase (East) Including East Penetration Area The main steam and feedwater pipe chase (east) is a reinforced concrete structure which houses and protects the main steam and feedwater piping. The overall dimensions are 127.58 feet long, 22.25 feet wide for most of the length, and 61.5 feet high. It is 41.25 feet wide at the widest point.

The east penetration area is a room located at the southern end of the pipe chase which houses the control panels for the hydrogen recombiner. See Figure 1.2-3 and Figure 1.2-4. j. Main Steam and Feedwater Pipe Chase (West) Including Mechanical Penetration Area and Personnel Hatch Area The main steam and feedwater pipe chase (west) is a reinforced concrete structure which houses and protects the main steam and feedwater piping. The pipe chase is 113.75 feet long, 20.0 feet wide for most of the length, and 61.5 feet high (overall); the pipe chase area is 59 feet high. It is 23.75 feet wide at the widest point. Located below the chase area is the mechanical penetration area which houses piping running between the Containment and the Primary Auxiliary Building. This region is partitioned into several smaller areas which include the radiation and nonradiation shield tunnels. The mechanical penetration area is 78.0 feet long, 34.71 feet wide at the widest point, and 37.5 feet high.

The personnel hatch area is an irregularly shaped reinforced concrete structure located outside the personnel hatch of the containment, for which it provides protection from missiles and illegal entry. It is connected to the west pipe chase and has approximate overall dimensions of 46.75 feet length, 37.25 feet width, and 29.25 feet height. See Figure 1.2-2, Figure 1.2-3, Figure 1.2-4, and Figure 1.2-5. k. Piping Tunnels The piping tunnels are underground reinforced concrete structures, rectangular in section, which are located throughout the plant. They house the piping systems found in the plant. See Figure 3.8-33.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 130 l. Pre-Action Valve Building The Pre-Action Valve Building is an irregularly shaped reinforced concrete building which contains the deluge valve for the Fire Protection System. It is approximately 34.83 feet long, 34.75 feet wide and 27.33 feet high in overall dimensions and is located on the east side of the Emergency Feedwater Building. See Figure 1.2-51. m. Primary Auxiliary Building Including Residual Heat Removal (RHR) Equipment Vault The Primary Auxiliary Building is a reinforced concrete structure. The major portion is approximately 79 feet wide by 145 feet long and extends from 13 feet and 46 feet below grade to 88 feet above grade. The equipment vault is attached to the structure and is 57 feet by 43 feet. It extends from 81 feet below grade to 5.5 feet above grade. The entire building is founded on rock or concrete fill.

The RHR equipment vault is subdivided into six compartments by continuous concrete walls as follows: two for containment spray pumps and heat exchangers, two for residual heat removal pumps and heat exchangers and two for access stairs. Plugs are provided in the reinforced concrete roof for removal of the heat exchangers. Located on the east and north sides of the RHR equipment vault are a personnel walkway and an electrical cable tray chase located above it. This structure is supported on fill concrete over rock and connected to the RHR equipment vault at its base, Elevation 20'

-8"; it extends to Elevation 53'

-0". The east portion is approximately 13.0 feet wide and 41.0 feet long; the north portion is approximately 10.0 feet wide and 38.0 feet long.

That portion of the building which is 79 feet by 145 feet, the Primary Auxiliary Building, has two intermediate reinforced concrete floors which support miscellaneous auxiliary nuclear equipment, such as heat exchangers, pumps, demineralizers, filters, tanks and ventilation equipment. Reinforced concrete walls and steel columns support the intermediate floors and reinforced concrete roof slab.

Below-grade reinforced concrete pipe tunnels connect the building to the Containment and Waste Processing Building. Monorail hoists are provided to handle materials and servicing of equipment.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 131 See Figure 1.2-9, Figure 1.2-10, Figure 1.2-11, and Figure 1.2-12 for plans and sectional elevations of the Primary Auxiliary Building and Figure 1.2-13 and Figure 1.2-14 for plans and sectional elevations of the RHR equipment vault, including the personnel walkway and electrical cable tray chase.

n. Safety-Related Electrical Duct Banks and Manholes Safety-related electrical duct banks are reinforced concrete structures, rectangular in cross section, which enclose cables running between various seismic Category I buildings other than the containment. Cross

-sectional dimensions vary from duct to duct and lengths are as required by the separation of the buildings being connected. The safety

-related electrical duct banks are completely below grade and principally supported on engineered fill. They are isolated from the buildings and manholes by flexible connections at the connection points.

Direction changes of the duct banks between buildings are accomplished by intermediate reinforced concrete manholes, which are generally rectangular in cross section. Overall dimensions vary depending on the size and number of duct banks entering the manhole.

See Figure 3.8-34 and Figure 3.8-35 for the duct banks and manholes respectively.

o. Service Water Cooling Tower Including Switchgear Rooms The Service Water Cooling Tower is a rectangular building approximately 300 feet x 54 feet in plan, extending 28 feet below grade, and projecting 57.5 feet above ground. It is located on the south side of the plant and is founded on rock. The cooling tower serves as an ultimate heat sink in the unlikely event that the cooling water tunnels are rendered inoperative. The cooling tower houses pumps, fans, water distribution system and nozzles.

The switchgear rooms are approximately 54 feet x 26 feet in plan and extend from Elevation 22.0 feet to Elevation 46.0 feet. They are located on the east and west ends of the cooling tower and are an integral part of the structure. The switchgear rooms house the switchgear, substation, and motor control center for the cooling tower. See Figure 1.2-56. Operational characteristics of the cooling tower are described in Subsection

9.2.5. SEABROOK

STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 132 p. Service Water Pumphouse The Service Water Pumphouse (SWPH) is located on the plant site east of the Containment Building and is founded on rock. It is attached to and shares a foundation with the Circulating Water Pumphouse (CWPH), a non

-Category I structure which is designed so that its loss or failure will not impair the Service Water Pumphouse or System.

The structure under the SWPH is a reinforced concrete basin and is approximately

91 feet wide by 74 feet long; it extends from the operating floor, 1 foot above grade, to 63 feet below grade. The Service Water Pumphouse itself is approximately 118 feet wide by 78 feet long and extends 28 feet above the operating floor to Elevation 49'

-0". The walls, roof and interior columns are reinforced concrete.

A service water electrical control room on the west end of the SWPH is a reinforced concrete structure founded on rock and is integrally connected to the SWPH. It is approximately 48 feet wide by 38 feet long and extends to 19.5 feet above grade.

The Circulating Water Pumphouse is approximately 119 feet wide by 123 feet long and extends 2 8 feet above the operating floor to Elevation 49'

-0". The building is non

-Category I and consists of a steel frame covered with metal siding, from the grade to the roof level, and roofing. A non

-Category I trash removal room on the north end is a steel framed structure with metal siding and is approximately 39 feet wide by 17 feet long extending to 14 feet above grade to Elevation 34'

-0". The structure under the CWPH is a reinforced concrete basin and is approximately 110 feet wide by 123 feet long; it extends from the operating floor to 63 feet below grade.

The basin under the SWPH is integrally connected to the basin under the CWPH by a common east

-west wall. This common basin, the SWPH and the electrical room are designed as a unit.

See Figure 1.2-46, Figure 1.2-47,and Figure 1.2-48 for plans and sectional elevations of the Service Water Pumphouse as well as the Circulating Water Pumphouse.

Operational characteristics of the Service Water System are described in Subsection 9.2.1, and operational characteristics of the Circulating Water System are described in Subsection 10.4.5.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 133 q. Tank Farm (Tunnels) Including Dikes and Foundation for Refueling Water Storage Tank (RWST) and Reactor Makeup Water Storage Tank (RMUWST)

The tank farm area consists of a reinforced concrete portion and structural steel framing portion. The reinforced concrete portion including the foundation, dike walls, pipe tunnels and pipe chases are structures associated with safety

-related systems and are designed as Seismic Category I. Structural steel framing portion which includes steel framing, concrete roofing and metal siding, is used to enclose the area above the tanks and to form the motor control center and switchgear room. The steel framing portion is designated nonseismic Category I and designed and constructed so that the safe shutdown earthquake (SSE) would not cause the steel framing portion to collapse upon any safety

-related structure, system or component within or surrounding the tank farm area.

The tunnels provide a passageway for piping which runs between the Primary Auxiliary Building and either the Service Water Pumphouse or the Service Water Cooling Tower.

The dikes are reinforced concrete walls surrounding the tanks and extending to Elevations 42'

-0" and 30'

-0" for the RWST dike and the RMUWST dike, respectively. Structural steel framing is used between the tops of the dikes and the roof framing, which is also structural steel.

Overall dimensions are 152 feet in length, approximately 66 feet in width and 92 feet in height. See Figure 1.2-23, Figure 1.2-24, Figure 1.2-25 and Figure 1.2-27. r. Waste Processing Building The Waste Processing Building is approximately 108 feet wide by 189 feet long and extends 51 feet below grade to 91 feet above grade.

The building consists of a reinforced concrete portion and a structural steel portion with a 4" nominal concrete slab.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 134 The reinforced concrete portion between column lines 1 to 2 and A to D, housing radioactive gaseous waste equipment and carbon delay bed, is designed as seismic Category I. To maintain integrity of this portion, the entire structure is designed as seismic Category I, except for siding, girts and roofing on steel at El.

111'-0. However, local effects of seismic loads from system supports is not accounted for, since they do not affect overall safety of the structure.

The safety

-related portion of the concrete between column lines 1 to 2 and A to D is designed to withstand tornado loads. The structural main frame above El. 53'-0" is designed to withstand tornado wind loads.

The building contains liquid and radioactive gaseous waste processing and solid waste systems. A trucking facility is provided along the south wall for shipping drums and containers.

See Figure 1.2-22, Figure 1.2-23, Figure 1.2-24, Figure 1.2-25, Figure 1.2-26, Figure 1.2-27, Figure 1.2-28, Figure 1.2-29 and Figure 1.2-30 for plans and sectional elevations of the Waste Processing Building.

s. Nonseismic Category I Structures Several nonseismic Category I structures were designed against collapse onto seismic Category I structures due to tornado wind and SSE loadings. These structures can be found on the site arrangement plan, Figure 1.2-1, and include the following:

- Turbine Building

- Nonessential Switchgear Building

- Tank Farm Area (Steel Framing Portion)

- Steam Generator Blowdown Recovery Building The Turbine Building is described in Subsection 1.2.2, and the tank farm area (steel framing portion) is described earlier in this subsection.

The Nonessential Switchgear Building is located on the north side of the Control Building and houses and protects the electrical equipment used to provide lighting for the plant.

The Steam Generator Blowdown Recovery Building is located on the south side of the Waste Processing Building and tank farm area, and houses the Steam Generator Recovery System.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 135 Specific design measures taken to protect the effected seismic Category I structures are given in Subsections 3.3.2 and 3.7.2 for tornado wind and SSE loadings, respectively.

t. Service Water Access Vault The Service Water Access Vault (SWAV) is located underground on the plant site north of the Cooling Towers. It is a reinforced

-concrete structure approximately 14 feet wide by 38 feet long; it extends from 15 feet to 29 feet below grade. A precast concrete manway extends from the top of the vault to approximately 2 feet below grade. The vault is founded on fill concrete and rock. The SWAV provides access to the underground safety

-related A and B Train 24" service water piping to permit entry into the piping for field weld joint refurbishment and joint inspection. See Figure 3.8-37. 3.8.4.2 Applicable Codes, Standards and Specifications Codes, standards and specifications listed in Subsection 3.8.3.2, as well as the following US

NRC regulatory guides, are also applicable to other seismic Category I structures:

Regulatory Guide No.

Title 1.76 (4/74)

Design Basis Tornado for Nuclear Power Plants 1.117 (4/78)

Tornado Design Classification 3.8.4.3 Loads and Loading Combinations The seismic Category I structures discussed herein were designed, as applicable, for all credible conditions of loadings, including normal loads, loads due to severe and extreme environmental conditions, abnormal loads and missile loads. These loads were determined in accordance with the applicable codes, including ACI 318

-71 and AISC

-69, and were considered in normal and unusual load combinations to assure that the structures remain within the limits prescribed in Subsection 3.8.4.5. a. Design Loads The definitions of the loads used in the design of the other seismic Category I structure include the following:

1. Normal Startup, Operational, and Shutdown Loads Normal loads are those loads encountered during normal plant operation, startup and shutdown. They include the following:

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 136 (a) Dead Loads (D) Dead loads are all permanent gravity loads including, but not limited to, concrete walls and slabs, structural framing, piping, cable and cable trays, permanent equipment and miscellaneous building items. Hydrostatic pressures of liquids are also included in this category.

(b) Live Loads (L) Live loads are all temporary gravity loads including but not limited to normal snow loads, conventionally distributed and concentrated floor loads, and movable equipment loads, such as cranes and hoists. Equipment operating loads and impact factors are the greater of those recommended by the manufacturer or the applicable building codes. Unusual snow load (L s), which is greater in magnitude than normal snow load, was also used where applicable. Lateral earth pressures due to soil backfill (H) were used where applicable. Three types of lateral earth pressure loading, active, at rest and passive, were considered, with pressures determined by acceptable theories of soil mechanics. (c) Operational Thermal Loads (T o) These are the thermal effects and loads occurring during normal operating or shutdown conditions, based on the most critical transient or steady

-state condition.

(d) Operational Pipe Reactions (R o) These are the pipe reactions occurring during normal operating or shutdown conditions, based on the most critical transient or steady-state condition.

2. Severe Environmental Loads Severe environmental loads are those loads which could infrequently be encountered during the plant life. The following loads are included in this category:

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 1 37 (a) Operating Basis Earthquakes (E o) These are the loads generated by the operating basis earthquake, which is the earthquake of an intensity of the same level as the highest intensity that has been experienced historically at the site.

Only the actual dead load and weights of fixed equipment were considered in evaluating the seismic response forces. The horizontal and vertical design response spectra for the OBE were derived by applying a factor of 0.5 to the response spectra given for the safe shutdown earthquake (SSE), which is described below.

The effects of two (2) orthogonal horizontal components and one (1) vertical component of earthquake were considered and combined by the squa re-root-of-the-sum-of-the-squares method. Lateral dynamic soil loads, including hydrodynamic loadings, due to the OBE (H e) were also used where applicable.

(b) Wind Load (W) Wind loads generated by the design wind specified for the plant site were considered. See Subsection 3.3.1 for a discussion of wind velocity and applied forces.

3. Extreme Environmental Loads Extreme environmental loads are those loads which result from postulated events which are credible, but highly improbable. The following loads are included in this category:

(a) Safe Shutdown Earthquake Loads (E s) These are the loads generated by the safe shutdown earthquake, which is the maximum potential earthquake that could occur in the vicinity of the plant, based on geological and historical investigations. Only the actual dead load and weights of fixed equipment were considered in evaluating the seismic response forces. The horizontal and vertical responses on the structures were developed from the response spectra given in Figure 2.5-38 and Figure 2.5-39, the development of which is described in Subsection 2.5.2.6. The effects of two (2) orthogonal earthquakes and one (1) vertical earthquake were considered and combined by the square

-root-of-t he-sum-of-the-squares method. Lateral dynamic soil loads, including hydrodynamic loadings, due to the SSE (H s) were also used where applicable.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 138 (b) Tornado Loads (W t) Wind loads generated by the design tornado specified for the plant site were considered along with the differential pressure loads due to rapid atmospheric pressure drop and the tornado

-generated missile effects.

See Subsection 3.3.2 for a discussion of the design basis tornado and its associated wind and pressure loadings. See Section 3.5 for a discussion of tornado

-generated missiles. 4. Abnormal Loads Abnormal loads are those loads generated by postulated high energy pipe ruptures, particularly a rupture in the Reactor Coolant System resulting in a loss-of-coolant accident (LOCA).

(a) Accident Pressure (P a) This is the design pressure load occurring within a structure due to the DBA. (b) Accident Temperature (T a) This includes the thermal effects and loads generated by the DBA including T

o. (c) Accident Pipe Reactions (R a) Pipe reaction loads due to thermal conditions generated by the postulated pipe break, including R o, were considered in the design.

(d) Pipe Break Loads (R r) These are local effects on the structures due to the postulated pipe break, as follows:

(1) R rr = load on the structure generated by the reaction of a ruptured high energy pipe during the postulated pipe break. The time-dependent nature of the load and the ability of the structure to deform beyond yield were considered in establishing the structural capacity necessary to resist the effects of R rr.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 139 (2) R rj = load on the structure generated by jet impingement of a ruptured high energy pipe during the postulated pipe break. In general, direct impingement of steam on a wall of a structure does not produce significant design loadings due to the distance between the wall and the break location.

Where a break is postulated to occur close enough to a wall to produce a critical loading, a shield, or deflector, is provided, and the loading is transferred to the pipe whip restraint to which the shield is attached.

(3) R rm = the load on the structure resulting from the impact of a ruptured high energy pipe during the postulated pipe break. Since all high energy lines are constrained by pipe restraints, loading of this nature is prevented.

5. Site-Related Loads (a) Flood Loads (F) These are the loads resulting from the design basis flood including uplift and wave runup. See Section 3.4. for further discussion.

(b) Missile Loads (M) Internal missile loads, other than those defined as R r, are considered including an appropriate dynamic load factor. See Section 3.5 for a detailed discussion of these missiles.

b. Load Combinations Various load combinations were considered in design to determine the strength requirements of the structure. Where varying loads occur, the combinations producing the most critical loading were used. The basic combinations considered in the design of each seismic Category I structure are given in Table 3.8-16. Two categories of loading conditions and criteria were used in the design of the

seismic Category I structures other than the containment, as described below:

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 140 1. Normal Load Conditions Normal load conditions are those encountered during testing and normal operation and are referred to in the standard review plan as service load conditions. They include dead load, live load and anticipated transients, loads occurring during normal startup and shutdown, and loads occurring during emergency shutdown of the nuclear steam supply, safety and auxiliary systems. Normal loading also includes the effect of an operating basis earthquake and normal wind load. Under each of these loading combinations the structures were designed so that stresses are within the elastic limits. Design and analysis procedures are presented in Subsection 3.8.4.4 and stress limitations are presented in Subsection 3.8.4.5. 2. Unusual Load Conditions Unusual load conditions are those resulting from combinations of accident, wind, tornado, earthquake, live and dead loads and are referred to in the standard review plan as factored load conditions.

For these loading combinations, the structures were designed to remain below their ultimate yield capacity such that deformations will be small and structural components will respond elastically. Design and analysis procedures are presented in Subsection 3.8.4.4 and stress limitations are presented in Subsection 3.8.4.5. c. Other Load Considerations

1. Creep Effects of concrete creep are negligible due to the low sustained concrete stresses associated with conventionally reinforced concrete structures and, therefore, were not a governing factor in design.
2. Stability The other seismic Category I structures were checked for overturning, sliding, and flotation using the load combinations of Table 3.8-16 with the exception that the coefficient for live load is zero. Buoyant forces were considered to decrease the dead load in computing both overturning and sliding.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 141 3.8.4.4 Design and Analysis Procedures

a. Design and Analysis Category I structures other than the containment are constructed of reinforced concrete with structural steel framing used to support vertical loading on floor slabs. Structural steel is also used to provide enclosure for some areas, as described in Subsection 3.8.4.1, and for other miscellaneous purposes.

Reinforced concrete structures consist of a system of walls and slabs generally to provide a continuous, integral framing system. Vertical forces are transferred to the foundation mats through the walls and structural steel and reinforced concrete columns. Lateral forces are transferred to the foundation mats primarily by the action of shear walls; some load is also transferred by means of flexural action of the walls, all of which are rigidly attached at the mat.

The Containment Enclosure Building, due to its cylindrical and hemispherical shape and relative dimensions, was analyzed as a three

-dimensional, thin

-shell structure. Boundary conditions were consistent with the support on rock and the lateral restraint provided by backfill concrete placed against the structure. Internal resultant forces and moments were determined by integration of the appropriate shell stresses through the thickness. Critical transverse shear force was derived by considering the variations in bending moments across the surface, in conjunction with the applied hydrostatic load (which produces additional local shear not reflected in the finite element analysis). Reinforcing was subsequently designed for these internal forces.

Columns are designed to resist other lateral loads, such as pipe loads or building displacements, in addition to those forces transmitted to the columns at floor levels. Steel columns are generally pin

-connected at the foundation mat, and reinforced concrete columns are rigidly attached. Structural steel framing for floor systems primarily consists of pin

-connected framing with some members being continuous. Rigidity is provided by the box

-like concrete walls and slabs.

Several computer programs were used for static analysis and are described in Append ix 3F. Table 3.8-17 contains a list of these programs and the respective structures for which they were used. The load combinations are given in Subsection 3.8.4.3.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 142 The idealization of each structure depended on its geometric configuration and applied loading. Both two and three

-dimensional analyses were performed. When the loads are considered to act principally in a given plane the floor diaphragms are rigid in their own planes; and when the rotations between adjacent frames are uncoupled, two

-dimensional analyses were used. Three

-dimensional analyses were used when other analyses were not suitable.

Detailed analyses of local areas of these structures, which either are subjected to local loads such as thermal hydrostatic and hydrodynamic loads, or have unusual geometry, such as numerous openings, were performed where required. Such areas included the north wall of the main steam and feedwater pipe chase (east) and the dividing wall in the Control and Diesel Generator Building. In these analyses, a representative area of a structure was isolated, and force and/or displacement boundary conditions, consistent with the overall behavior of the structure, were imposed on the substructure in addition to the appropriate loading.

Structural systems of the buildings were designed for dead load, live load, and lateral loads, such as those loads produced by wind, tornado, and earthquake. All structural elements were designed to resist the effects of internally generated missiles, where applicable. Tornado loads consist of applied pressure and missile impact, for which all seismic Category I structures are designed except as indicated below.

The safety

-related electrical duct banks and manholes, the mechanical penetration area and the pipe tunnel were not designed for any tornado loads. The duct banks and manholes are covered by a nominal 5 feet of backfill, minimum, which protects them from tornado loads. The mechanical penetration area is located beneath the west pipe chase, and the pipe tunnel is covered by approximately 15 feet of backfill. The manhole covers, however, were designed for tornado-generated missiles.

The Service Water Cooling Tower and the Containment Enclosure Building were designed for tornado pressure but not for tornado

-generated missiles. Since the circulating water tunnels will be in operation during a tornado instead of the cooling tower and since the containment was designed for tornado

-generated missiles, these structures need not be designed for missiles.

The Pre-Action Valve Building is designed for tornado

-generated missiles and tornado depressurization loading (Elevation 27'

-0" slab). The elevation 27'

-0" slab is a tornado depressurization barrier for safety

-related equipment and components located in the electrical tunnels.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 143 In the Waste Processing Building only those areas housing radioactive gaseous waste equipment were designed for tornado loads; the other areas do not contain safety-related equipment. Procedures by which the structures were checked for missile loads, tornado-generated as well as internally generated, are described in Section 3.5. Determination of pressures on structures due to tornado is described in Subsection

3.3.2. Pressure

loadings from wind are described in Subsection 3.3.1. The seismic analysis of seismic Category I structures is described in Subsection 3.7.2. All cranes in these structures are furnished with hold

-down devices to ensure that they are not dislodged by earthquake forces. Monorails, by nature of their support mechanisms, cannot be dislodged by earthquake forces.

Using methods outlined in TID

-7024, "Nuclear Reactors and Earthquakes," the effects of hydrodynamic forces were included in the seismic analyses of the Service Water Cooling Tower. Also using methods outlined in TID

-7024, the weight of constrained water and sloshing effects of water in motion were included as equivalent static loads in the final design of the Service Water Cooling Tower, Service Water Pumphouse and Fuel Storage Building.

Reinforced concrete design of Catego ry I structures was in accordance with the strength design procedures of the ACI 318-71 code, except as indicated in Subsection 3.8.4.5. Structural steel design was in accordance with the provisions of the AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings (1969 Edition). Refer to Subsection 1.8 (Regulatory Guide 1.142) for a statement concerning compliance with ACI

-349. b. Material Properties Material properties were selected from the normal range of values to produce a conservative design. See Subsection 3.8.1.4 for a detailed discussion of the influence of material properties on design and analysis.

c. Computer Programs The computer programs used in analysis and design of other Category I structures are listed in Table 3.8-17. These programs are further described in Appendix 3F.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 144 3.8.4.5 Structural Acceptance Criteria The basis for the acceptance criteria is the ACI 318 and AISC Codes. However, under the action of seismic or wind loadings, in accordance with the standard review plan (Section II.5), the 33 percent increase in allowable stresses was not permitted.

a. Normal Load Conditions Structures were proportioned to remain within the elastic limits under all normal loading conditions described in Subsecti on 3.8.4.3. Reinforced concrete structures were designed in accordance with ACI

-318 Strength Method, which insures flexural ductility by limiting reinforcing steel percentages and stresses.

Structural and miscellaneous steels were designed in accordance with AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings, Part

1. b. Unusual Load Conditions Structures were proportioned to maintain elastic behavior under all unusual load conditions shown in Subsection 3.8.4.3. The upper bound of elastic behavior was taken as the yield strength capacity of the load carrying components. The yield strength of structural and reinforcing steel was taken as the minimum guaranteed yield stress as given in the appropriate ASTM Specifications. Reinforced concrete structures were designed in accordance with ACI

-318 Building Code. Member yield strength was considered to be the strength capacity calculated by the ACI Code. Structural and miscellaneous steels were designed in accordance with Part 1 of AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings.

c. Deformations Since each of the structures was designed to be in the small deformation, elastic range, no gross deformations will occur that will cause significant contact with other structures or pieces of equipment. All deformations, however, were evaluated considering the relationship of the subject component to both adjacent and supporting structures and equipment.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 145 d. Additional Stress

-Strain Considerations Stress-strain limits and design parameters were applied to the design of each applicable element for the specific values identified with each loading combination and design condition identified in Subsection 3.8.4.3. No special allowance has been made for variation of material properties over the life of the structure, beyond that which is taken into account in establishing allowable stresses, strains, capacity reduction factors, concrete protection of reinforcing, and crack control as outlined in the referenced ACI and AISC codes.

Additional corrosion protection is provided to concrete structures by means of waterproofing for parts of the structure below grade and by painting, coating or installing of liners for structural concrete tanks (such as the spent fuel pool in the Fuel Storage Building).

Variations in stress and strain, due to scheduled plant shutdowns and startups, have negligible effect on the overall structural behavior because of the small variation in the average structure temperature and loading. Since the designs of the structures were governed by extreme, infrequently occurring loadings, such as tornadoes and earthquakes, and normal cyclical changes in stress levels are comparatively small, no reduction in the margin of safety will occur over the life of the plant.

All connections and joints were designed to transfer all design forces (shear, tension and compression) and moments with a safety margin and degree of conservatism that is required by the applicable code.

e. Stability Acceptance criteria for stability are given in Subsection 3.8.5.5.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 146 3.8.4.6 Materials, Quality Control and Special Construction Techniques The primary materials of construction are concrete, reinforcing steel and structural steel (rolled shapes and plates).

Descriptions of the materials and basic quality control procedures are discussed in Subsection 3.8.3.6. Alkali-aggregate reactions (AAR) occur over time in hardened concrete between the alkali hydroxides in the pore solution of concrete and certain minerals found in some aggregates.

Alkali Silica Reaction (ASR) is the predominant type of AAR. It involves a chemical reaction between alkalis in the cement paste (Portland cement) and reactive forms of silica in the aggregates.

This reaction is depend ent on several factors including the amount and form of reactive material in the aggregate (e.g.

, reactive forms of quartz), the amount of alkali in the cement (more alkali

- faster reaction), temperature (higher temperature

- higher reaction rate), and moisture content.

The reaction forms an expansive gel in the affected material. As the reaction progresses and the gels expand, micro

-cracks are formed in the aggregate extending into the cement paste.

The main observable effect of ASR affected structures is expansion and cracking due to gel formation.

As expansions increase, visible cracks begin to form on the exposed surfaces. These cracks are often in a characteristic pattern cracking and may also have signs of ASR gel material.

While very reactive aggregates can cause rapid expansion rates that manifest in visible cracks and measurable expansion rates in a few years, ASTM testing for reactive aggregates and specification of low alkali cement ha ve been somewhat effective in preventing ASR in these time frames.

Slow reacting aggregates may not manifest for decades.

The concrete constituents used at Seabrook would not be expected to be susceptible to ASR since: 1. the coarse aggregate is largely igneous rock that was routinely tested during construction and passed petrographic examinations and expansive reaction tests that normally detect alkali

-silica reaction; and

2. low-alkali Portland cement was abundantly used.

Aggregates routinely passed ASTM reactivity and expansion tests per C227 and C289.

Petrographic examinations of aggregates per C295 were performed but did not detect presence of reactive aggregates. In retrospect, the testing standards in place at the time of original construction may not accurately predict late or slow reactive aggregates. Empirical evidence at Seabrook suggests the coarse aggregates are the slow reactive type.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 147 In June 2010, concrete core s were removed for examination and testing from the walls of the lower electrical tunnel in the Control Building, as part of preparations for license renewal inspections. The purpose was to evaluate potential concrete aging effects in below grade areas of the plant that had been subjected to historical groundwater wetting of the concrete. In general the removed cores showed the expected quality materials and placements from original construction. There were no obvious visual signs of aging distress or concrete degradation.

Petrographic examinations were performed which involved sectioning and polishing the core samples and analysis under magnification by a qualified professional petrographer.

This analysis revealed micro cracks and other features indicative of Alkali Silica Reaction (ASR). Materials testing of the removed cores also resulted in lower than expected mechanical properties consistent with low levels of ASR. The impact of ASR in the material strength testing of removed cores is not indicative of actual in situ performance and cannot be directly correlated to actual structural impact. Once removed from the structural context (e.g., reinforcement or confining loads) the behavior of the cores no longer reflects that of the confined structure.

Additional concrete core sampling has been performed to determine the extent of condition both from the perspective of additional areas that might be affected by ASR and also the extent of ASR degradation within a given area and control areas (non

-wetted adjacent areas). Subject Matter Experts from around the country were consulted and a specific monitoring and action plan for ASR was added to the Structural Monitoring Program. Engineering evaluations that were performed and documented in foreign print 100716, (Subsection 3.8.6, Ref. 6) established that although the concrete can be considered degraded, the structures and embedded concrete anchors are capable of performing all required design basis functions.

ASR is considered to be a degraded nonconforming condition pursuant to Regulatory Issue Summary (RIS) 2005

-20. An operability determination was performed which established reasonable assurances that the structures and embedded/drilled

-in concrete anchors are capable of performing all required design basis functions.

Design basis calculations will be reconciled to account for ASR following completion of the actions delineated in the ASR corrective action plan.

3.8.4.7 Testing and In

-Service Surveillance Requirements Normal quality control testing is discussed in Subsection 3.8.3.6. A general visual inspection of the exposed accessible interior and exterior surfaces of the Containment Enclosure Building will be periodically conducted as discussed in Subsection 6.2.6.1.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 148 3.8.5 Foundations The following sections discuss the physical descriptions of the foundations, applicable codes, standards and specifications, loads and load combinations, design and analysis procedures, structural acceptance criteria, materials, quality control and special construction techniques, and testing and in

-service inspection requirements for the foundations of seismic Category I structures.

3.8.5.1 Descriptions of the Foundations The locations and relationships of the various seismic Category I foundations are shown on the plot plan, Figure 1.2-1. Details of individual foundations, including type and dimensions, are given in Table 3.8-15. This table also contains a list of figures showing plans and profiles of the foundations.

Foundations for seismic Category I buildings are conventionally reinforced concrete mats of varying thicknesses supported on sound bedrock or on fill concrete extending to sound bedrock. The walls of the Containment Enclosure Building extend to a spread footing, 10'

-3" wide by 10' deep, which carries the load from the walls to sound rock. This footing is not continuous, having openings for the pipe chases and electrical tunnels below the Emergency Feedwater Pumphouse.

The bottoms of this footing and most mats are embedded in the rock in order to transfer horizontal shear forces. If the bottom of a foundation does not extend to sound rock, fill concrete was placed from the sound bedrock to the elevation of the underside of the structure.

The only exceptions to the above criteria are five safety

-related electrical manholes, certain sections of service water pipes, and most of the 8350 feet of safety

-related electrical duct banks which are supported on engineered backfill consisting of offsite borrow or tunnel cuttings (see Subsections 2.5.4.5c and 3.7(B).1.4). Compaction requirements for this engineered backfill ensure firm support which, along with flexible couplings between the manholes and ducts, prevent any seismic response in one of these components from inducing a response in an adjacent component or structure.

Since all other seismic Category I structures are founded on sound rock, or on fill concrete over sound rock, they are inherently isolated from each other. There is no connecting soil medium which could foster seismic coupling.

Similarly, the foundations of nonseismic Category I structures, which are also founded on rock, do not have seismic coupling with the foundations of seismic Category I structures, except for the Circulating Water Pumphouse. This is a non

-Category I structure which is attached to the Service Water Pumphouse, shares a common foundation it, and it is designed as a unit. Seismic isolation of the buildings above the point of fixity is ensured by means of isolation joints; these joints are provided between adjacent buildings above the point of fixity.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 149 The point of fixity is the elevation on the building to which rock extends or fill concrete was placed. There is no correlation between this point and the natural rock elevation; it may be lower than natural rock or higher if fill concrete was used to increase the embedment. Within the region of a building that is fixed, the walls and foundation are rigid and accelerate along with the rock. There are no unique features used for foundations.

The load transfer for the foundations is as follows:

a. Load is applied to the wall of a seismic Category I structure.
b. The wall transfers the load to the foundation (mat or footing). c. The foundation transfers the load to the rock, fill concrete, or engineered fill.

A more detailed description of the mechanism of load transfer is given in Subsection 3.8.5.4. For a detail of the typical reinforcing pattern at the junction of a reinforced concrete wall and its foundation, see Figure 3.8-36. Piles are not used under seismic Category I structures. The structures transmit load directly to rock by contact or through fill concrete or engineered fill.

3.8.5.2 Applicable Codes, Standards and Specifications For applicable codes, standards and specifications, see Subsections 3.8.1.2 and 3.8.4.2 for the containment structure foundation and for other seismic Category I structure foundations, respectively.

3.8.5.3 Loads and Load Combinations For the containment structure foundation, see Subsection 3.8.1.3 for loads, load combinations, load factors and the design approach used with the load combinations and load factors.

For other seismic Category I structure foundations, see Subsection 3.8.4.3 for loads, load combinations, load factors and the design approach used with the load combinations and load factors. Foundations and structures are checked for sliding and overturning due to earthquakes, winds, and tornadoes and for flotation due to floods and high water table using load combinations described in Subsections 3.8.1.3 and 3.8.4.3 for the containment structure and the other seismic Category I structures, respectively.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 150 Lateral earth pressures are considered as applicable. Nonrigid walls are designed for active earth pressure under static conditions and for an equivalent static earth pressure based on a coefficient of dynamic earth pressure that dampens with depth under seismic conditions. Rigid walls are designed for earth pressure at rest under static conditions and for an equivalent static earth pressure based on a constant coefficient of dynamic earth pressure under seismic conditions.

Lateral pressure diagrams for these conditions are shown in Figure 2.5-52 and Figure 2.5-53 and discussed in Subsection 2.5.4.11. 3.8.5.4 Design d Analysis Procedures The foundations of seismic Category I structures are analyzed and designed in accordance with Subsections 5.2 and 3.8.5.5 to determine maximum stresses in reinforcing and concrete, using the load combinations discussed in Subsection 3.8.5.3. a. Boundary Conditions and Expected Behavior Most foundations of seismic Category I structures are founded directly on sound rock or on fill concrete; five safety

-related electrical manholes and most of the safety-related electrical duct banks are supported on engineered fill. The entire length of the duct banks, however, is designed with the assumption of support on engineered fill.

Design and analysis, including idealization and boundary conditions, for the circular base mat of the containment structure are described in Subsection 3.8.1.4. The base mat is designed to sustain all credible loads resulting from the containment and internal structures.

Design procedures for all structures insured that foundation mats and footings were sized to limit bearing pressure on the rock to 60 tons/square foot on horizontal surfaces and 10 tons/square foot in weathered zone and 60 tons/square foot below weathered zone on vertical surfaces. These bearing pressures were established on the basis of results of tests of unconfined compressive strength with a factor of safety on the order of

10. Foundations for those structures supported on engineered fill are designed to limit bearing pressures to the allowable limits given in Subsection 2.5.4.5. Each building has an individual foundation; no common foundations are employed except for the Service and Circulating Water Pumphouse. In a few additional cases, as noted in Subsection 3.8.4.1, a structure is partitioned into a

dual-function structure which is designed as a unit. Also, adjacent foundations such as those for the containment structure and the Enclosure Building, even though they are not continuous, may transfer horizontal forces through bearing.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 151 All loads are transferred to the foundation through the walls, or in the cases of some equipment, through the supporting pedestals. Reinforcing continues from the wall into the foundation mat and is embedded in the mat and wall by hooks, bends or straight extensions of sufficient length to develop the design strength of the bars. Thus, forces are adequately transferred from the wall to the foundation.

Soil-structure interaction is not applicable for rock

-supported seismic Category I structures. Thus, for these structures, the foundation mats are designed as flat plates supported on rigid non

-yielding foundation media.

In the case of electrical duct banks in engineered fill, the ability of the elastic fill to impose a seismically induced strain on the duct bank was considered. This is further described in Subsection 3.7(B).2. For safety

-related electrical manholes which are supported on engineered fill, amplification of the ground motion is considered. The slab, however, is essentially rigid, and bearing pressures are very low. (See Subsection 3.7(B).2.)

Behavior of the foundation mats when subjected to the load combinations given in Subsection 3.8.5.3 is within the acceptance criteria of Subsection 3.8.5.5. b. Vertical Loads, Lateral Loads and Overturning Moments Vertical loads are carried by direct bearing on the rock. Overturning moments and bearing pressures on the rock (including maximum toe pressure) were investigated in accordance with the safety factors discussed in Subsections 3.8.1.3 and 3.8.4.3 for the containment foundations and other seismic Category I structure foundations, respectively.

Base shears and torsional moments are transferred to rock by one or both of the following:

1. Embedment in rock which provides resistance to shearing loads through direct bearing 2. Friction between the bottom of the foundation and the top of the rock (the coefficient of friction takes into consideration the reduced shear resistance due to the presence of the waterproofing membrane).

In the containment structure, the walls of the reactor cavity bear against the rock through fill concrete to provide additional resistance for transfer of base shears.

The walls of the cavity are designed to carry these loads.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 152 Conventional hand calculations were employed for the design of the foundations and were augmented, for some of the larger foundations, by finite element techniques to determine uplift patterns.

This uplift was determined by static methods using forces which were determined in the dynamic seismic analysis. The mat was modeled by a series of flat plate bending elements which include stiffening effects of the walls or shell. An uplift pattern was assumed, loads applied, and final displacements checked. When the calculated displacements were in agreement with the assumed uplift pattern, the analysis for that load combination was then completed by calculating shears and bending moments.

Methods of determining the overturning moments which were used in the uplift analysis include all three components of the earthquake and are discussed in Subsection 3.7(B).2. Design factors of safety against sliding, overturning and buoyancy are defined in Subsections 3.8.1.3 and 3.8.4.3 for the containment structure and other seismic Category I structures, respectively.

c. Computer Programs See Appendix 3F for a description of the programs. See Table 3.8-17 for a list of structures and the computer programs which were used for them.

3.8.5.5 Structural Acceptance Criteria The acceptance criteria relating to stress, strain, gross deformation and shear loads are described in Subsections 3.8.1.5 and 3.8.4.5 for the containment and other seismic Category I structure foundations, respectively.

Safety factors for buoyancy, sliding, and overturning are as follows:

Factor of Safety Load Overturning Sliding Flotation Service/Normal load combinations 1.5 1.5 -- Factored/Unusual load combinations 1.1 1.1 -- Dead load and design basis flood load

-- -- 1.5 SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT AND SYSTEMS Design of Category I Structures Revision 16 Section 3.8 Page 153 3.8.5.6 Materials, Quality Control and Special Construction Techniques The primary materials of construction are concrete and reinforcing steel. Their descriptions and basic quality control procedures are discussed in Subsections 3.8.1.6 and 3.8.4.6 for the containment foundation and other seismic Category I structure foundations, respectively.

Engineered fill, fill concrete and backfill concrete are described in Subsection 2.5.4.5. There are no special construction techniques.

3.8.5.7 Testing and In

-Service Surveillance Requirements The ability of the containment foundation to resist 1.15 times the design pressure is demonstrated during the structural integrity test as described in Subsection 3.8.1.7. For other seismic Category I structure foundations, no preoperational or in

-service surveillance is required. Structures which are founded on sound rock or on fill concrete over sound rock do not have any potential areas of settlement or displacement which should be monitored. Similarly, gradation requirements, compaction criteria and compaction tests for engineered fill ensure a foundation material which will support the design loads with negligible settlement. Piles were not used. For these reasons there are no potential settlements or displacements which should be monitored for any foundation.

3.8.6 References

1. Wilson, E.L., "Structural Analysis of Axisymmetric Solids," AIAAJ. Vol

. 3, No. 12 (1965) pp. 163

-182. 2. Wilson, E.L., "A Digital Computer Program for the Finite Element Analysis of Axisymmetric Solids with Orthotropic, Nonlinear Material Properties," November

1969. 3. Duchon, N.B., "Analysis of Reinforced Concrete Membrane Subject to Tension and Shear," ACI Journal, September, 1972, pp. 578

-583. 4. Wilson, E.L. and Nicholl, R.E., "Application of the Finite Element Method to Heat Conduction Analysis," Journal of Nuclear Engineering and Design, Vol. 4, 1966. 5. Alexandria, S. C., Effects of Irradiation of Concrete, Final Results, Research Reactor Division UKAEA, Harwell, AERE

-R-4490, December 1963.

6. Foreign Print 100716, Impact of ASR on Structures

.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 14 Section 3.9(B)

Page 1 3.9(B) MECHANICAL SYSTEMS AND COMPONENTS 3.9(B).1 Special Topics for Mechanical Components 3.9(B).1.1 Design Transients Refer to Subsection 3.9(N).1.1. 3.9(B).1.2 Computer Programs Used in Analyses Computer programs used in dynamic and static analyses of seismic Category I code and noncode items are described below for their theories, assumptions, references and applications.

a. ADLPIPE The stresses and loads in piping systems due to thermal expansion, dead weight, and seismic loadings are calculated by using the ADLPIPE program. ADLPIPE is a computer program developed by Arthur D. Little, Inc., of Cambridge, Massachusetts, and is used for static and dynamic elastic analyses of complex piping systems. The piping is mode led as a series of sections that lie between network points. A section is com posed of a sequence of straight and/or curved pipes, and each pipe may have common or different loads and physical properties. The network points may be free, partially or fully restrained, and have specified displacements that represent thermal anchor displacements or seismic anchor motion. Intermediate springs to ground or joining other members may be placed within the section to represent spring hangers, pipe bellows, skew and guided restraints, supports and equipment. The matrix displacement method is used throughout the program, and transfer matrix techniques are used to reduce the size of the system stiffness matrix. The ADLPIPE computer program provides the ability to compute a summary stress report which can be used for direct comparison with the criteria of ASME Section III. Verification test cases for ADLPIPE have been performed by UE&C's computer service vendor. Test problems have been executed and documented which agree with the program author's verification documents. Acceptable verification criteria are: (1) hand calculations, (2) known theoretical and experimental solutions and (3) solutions from other verified computer programs.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 14 Section 3.9(B)

Page 2 b. MRI/STARDYNE This computer program uses the finite-element method for the static and dynamic analysis of two- and three-dimensional linear elastic structural models, subjected to any arbitrary static or dynamic loading. Static results and dynamic responses can be presented as structural deformations, and/or internal member loads/stresses. Using either the direct integration or the normal mode techniques, dynamic response analysis can be performed for a wide range of loading conditions, including transient, steady-state harmonic, random and shock spectra excitation. The MRI/STARDYNE code is used in the analysis of seismic Category I cable tray and conduit systems. Verification for the STARDYNE program was accomplished in the same manner as that used for ADLPIPE.

c. ANSYS ANSYS is a large-scale general purpose computer program. The matrix displacement method of analysis based upon finite element idealization is used throughout the program. Analytical capabilities include static and dynamic, small and large deformations, steady-state and tr ansient heat transfer and steady-state fluid flow. Loading may be forces, displacements, pressures and temperatures. Verification for the ANSYS program was accomplished in the same manner as that used for ADLPIPE.
d. ICES/STRUDL-II This computer program provides the ability to perform static and dynamic analysis for framed structures and three-dimensional solid structures. The matrix displacement method of analysis based on finite element idealization is used throughout the program. Verification for the ICES/STRUDL-II program was accomplished in the same manner as that used for ADLPIPE.

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Page 3 e. ADLPIPE-2 A static and dynamic pipe design and stress analysis program developed by Arthur D. Little, Inc., and modified by UE&C, the program provides elastic analyses of piping systems in accordance with the requirements of ASME Section III - Nuclear Power Plant Components, ANSI B31.1 - Power Piping and USNRC Regulatory Guide 1.92. Analytical capabilities of the program include: Deadweight Thermal Anchor movements Pressure External forces Seismic response spectra including closely spaced modes (Regulatory Guide 1.92). The validity of ADLPIPE-2 results has been verified by comparing results of identical problems analyzed by other verified computer codes or hand calculations. Computations of forces, moments and displacements have been verified against the original ADLPIPE program, the validity of which is supported by Arthur D. Little, Inc., in documentation of ADLPIPE. The ADLPIPE-2 program was further verified in 1979 by a series of benchmark piping problems presented by the NRC in BNL-NUREG-21241-R2. Supplemental verification has been achieved by demonstrating exceptionally good agreement between ADLPIPE-2 results and the results published for Problem #4 in NUREG/CR-1677. Comparisons of elemental forces, reaction forces, moments, displacements, eigenvalues, eigenvectors, and stress computations were made. Computations for determining the results of a three component earthquake using square-root-of-the-sum-of-the-squares (SRSS) summations, closely spaced modes

and stress intensification factors have been verified by hand calculations. All of the above verification efforts have been fully documented and are available for review. This program has been used for analysis of ASME Section III, Class 2 and 3, and ANSI B31.1 piping systems.

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Page 4 f. IMAPS The IMAPS (Interactive Mode Analysis of Piping and Structures) computer code performs static linear elastic analyses of three-dimensional multisupported simplified piping systems of framed structures. Analytical capabilities of the program include: Deadweight Thermal Anchor movements External forces and moments Seismic static. The validity of the results for IMAPS has been verified by comparison of results for identical problems performed on the ADLPIPE-2 program. In addition, IMAPS results have been compared with benchmark piping problems from USNRC (BNL-NUREG-21241-R2) with excellent comparison of results. This program has been used for analyses of small ASME Section III, Class 2 and 3, and ANSI B31.1 piping.

g. SPHNOZ/CYLNOZ This program computes local stresses in spherical and cylindrical shells due to external loads. The program is based on the graphical results of the analytical work by Prof. Bijlaard, published in We lding Research Council Bulletin No. 107, Rev. 2, July 1970, Franklin Institute Research Laboratories, Philadelphia, Pa. The program results have been verified by comparison with hand calculations based on WRC Bulletin No. 107. This program has been used as an alternative method to certify the design adequacy of branch piping and nozzles.

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Page 5 h. WATER Program WATER determines the transient forces which result from rapid opening of a safety/relief valve in a piping system. Initially, water is upstream from the valve and air is downstream from the valve. Water discharges through the valve, and downstream conditions are such that flashing does not occur. Program output consists of total force vs. time history at each elbow in the downstream piping system, or, equivalently, the net force/time history on each piping leg. The program is based upon in-house formulation using rigid water column theory. Verification consists of demonstrating that program results and those obtained from a hand calculation agree for identical input data. Additionally, fluid output parameters (acceleration and velocity) from program WATER agrees with results from program RELAP. Program RELAP is in the public domain.

i. VALCLO Program VALCLO determines the maximum force developed due to closing a valve with prescribed closing characteristics. Prior to valve closing, steam is flowing in the system. The maximum force is determined based upon formulas presented in "Steam Hammer in Turbine Piping Systems," Coccio, C.L.; ASME Publication 66-WA/FE-32, 1968. Verification consists of demonstrating that computer results and results obtained from a hand calculation agree for identical input data.
j. ELBFOR Program ELBFOR is a wave super-position program. It is run as a post processor to program VALCLO; however, ELBFOR coul d be used as a post processor to any source calculation which would establish the maximum force in the system.

Using the maximum force, ELBFOR establishes a ramp forcing function over the valve operating time. Considering system geometry, the ramp forcing function is propagated through the system at sonic velocity to establish the total force/time history for each elbow in the system, or, equivalently, the net force/time history for each piping leg in the system. Verification of ELBFOR consists of demonstrating that computer results and results obtained from a hand calculation agree for identical input data.

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Page 6 k. MADIS Program MADIS determines the transient forces which result from rapid opening of a safety/relief valve in a piping system. Initially, superheated or saturated steam is upstream from the valve and air is downstream from the valve. The program output consists of the total force/time history at each elbow in the downstream piping system or, equivalently, the net force/time history on each piping leg. The program is based upon formul as presented in "Analysis of Safety Valve Discharging into Closed Piping System;" Luk, C.H.; Structural Design of

Nuclear Plant Facilities, No. 1; Specialty Conference held at Chicago, Ill. on December 17-18, 1973. Verification of the program consists of demonstrating that computer results and those obtained from hand calculations agree for identical input data.

l. FLEXPLT FLEXPLT is an interactive time-sharing computer program used for the analysis of flexible baseplates. These baseplates have concrete expansion anchors as the means of attachment to the concrete. A variety of static load combinations may be applied through a single structural attachment. FLEXPLT has the capability of analyzi ng the baseplate with the structural attachment centrally located on the plate and the anchors within a given window, or it can analyze the baseplate with the anchors at any fixed location and the structural attachment off the centerline of the baseplate. FLEXPLT has been verified by comparison of results for identical problems using the PSBASEPLATE computer program and hand calculations. m. GT STRUDL GT STRUDL is a large-scale general purpose computer program. The matrix displacement method of analysis based upon finite element idealization is used throughout the program. GT STRUDL has the ability to perform static and dynamic analysis for framed structures and three-dimensional solid structures. GT STRUDL is used in the analysis and design of nuclear and nonnuclear linear type pipe supports and seismic Category I duct supports. The validity of results for GT STRUDL has been verified by comparison of results for identical problems performed using hand calculations.

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Page 7 n. PSBASEPLATE PSBASEPLATE is a pre- and postprocessing computer program which interfaces with STARDYNE. PSBASEPLATE is used fo r analyzing flexible baseplates on a geometrically nonlinear foundation. PSB ASEPLATE can generate the finite element model of a single structural attachment to a baseplate fastened to ground by concrete expansion anchor. PSBASEPLATE processes the output produ ced by STARDYNE to aid the user in interpreting the results. The validity of results for PSBASEPLATE has been verified by comparison of results for identical problems using other recognized computer codes. o. BASEPLATE II BASEPLATE II is a pre- and postprocessing computer program which interfaces with STARDYNE. BASEPLATE II is used fo r analyzing flexible baseplates on a geometrically nonlinear foundation. BASEPLATE II can generate the finite element model of either the plate-bolt-structure assembly, or a model including a framed structure attached to the baseplate. BASEPLATE II processes the output produ ced by STARDYNE to aid the user in interpreting the results. The validity of results for BASEPLATE II has been verified by comparison of results for identical problems using other recognized computer codes. p. McAuto STRUDL McAuto STRUDL is a large-scale general purpose computer program. The matrix displacement method of analysis based upon finite element idealization is used throughout the program. McAuto STRUDL has the ability to perform static and dynamic analysis for framed structures and three-dimensional solid

structures. McAuto STRUDL has extensive postp rocessing capabilities, including an ASME Section III, Subsection NF code check for linear type supports. Also, McAuto STRUDL has the capability to design a limited class of fillet welds in butt joint connections.

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Page 8 McAuto STRUDL is used in the anal ysis and design of nuclear and nonnuclear linear type pipe supports. The validity of results for McAuto STRUDL has been verified by comparison of results for identical problems performed using hand calculations or other recognized computer codes such as NASTRAN and ANSYS.

q. STFPSG STFPSG is an interactive, time-sharing computer program using closed form solutions to calculate the structural stiffness or moment of inertia of a limited class of structural frames and beams. STFPSG has been verified by comparing the computer-generated values with hand-calculated values.
r. PSGPREP PSGPREP is an interactive, time-sharing computer program, which will prepare input data for a McAuto STRUDL analysis from user keyboard input. PSGPREP is designed to produce input suitable for an analysis of a pipe support. However, PSGPREP can be used to handle analyses of other structural frames. PSGPREP has been verified by comparing its results to manually generated input. s. PSGWARP PSGWARP is an interactive, time-sharing computer program which calculates the stresses in beams due to torsional effects. The closed form solutions employed in PSGWARP were extracted from the Bethlehem Steel publication "Torsion Analysis of Rolled Steel Sections." PSGWARP will also, by superposition, combine bending and shear stresses with torsional stresses and compare them with ASME Code allowables (Subsection NF). PSGWARP was verified by comparing results with those from identical input supplied to the McAuto STRUDL Program.

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Page 9 t. CHADS CHADS is an interactive, time-sharing program that enables a user to create a pipe support mathematical model including loading conditions, member properties, material properties, and code checking data to generate input for a GTSTRUDL analysis. The user can also employ the same geometry data to create a complete detail drawing of th e support and then use GTSTRUDL and WELDDA to check member stress and weld design. CHADS also uses the GTSTRUDL output to prepare input for BASEPLATE II, if it is desired to analyze an anchorage c onsisting of concrete expansion bolts and plates. Verification of CHADS was accomplished in the same manner as for ADLPIPE. u. WELDDA WELDDA is a postprocessing program to the GTSTRUDL computer code for designing and analyzing welded connections between rolled structural members. WELDDA uses a closed form solution base d on analyzing the weld pattern as a

line to calculate weld stress and/or size. WELDDA then performs a code check to either the AISC or ASME Codes. Verification of WELDDA was accomplished in the same manner as for ADLPIPE. v. DIS (Design Information System) DIS is an interactive piping data management system developed by Arthur D. Little, Inc., for piping engineering. Equipment, pipe, and fabrication

specifications are stored in a data base. From these specifications, a piping analytical isometric can be developed. DIS can also be used to generate a bill of material and the ADLPIPE mathematical model. The user also controls printing

of various output reports. DIS was verified in the same manner as ADLPIPE.

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Page 10 w. NUPIPE II The stresses and loads in piping systems due to thermal expansion, dead weight, and seismic loadings are calculated by using the NUPIPE II program. NUPIPE II is a computer program de veloped by Quadrex Corporation of Campbell, California, and is used for static and dynamic elastic analyses of complex piping systems. The matrix displacement method of analysis based upon finite element idealization is used throughout the program. The piping is modeled as individual elements connected by node points. The node points may be free, fully or partially restrained, have specified displacements, or, connected to ground through an element with a finite stiffness to represent pipe supports or equipment. Spring stiffnesses may be used to connect elements to represent pipe bellows or

ball joints. Verification test cases for NUPIPE II have been performed by UE&C's computer service vendor. Test problems have been executed and documented and agree with the program author's verification documents. Acceptable verification criteria are: (1) hand calculations, (2) known theoretical and experimental solutions and (3) solutions from other verified computer programs. Verification for NUPIPE II was accomplished in the same manner as that used for ADLPIPE. x. CONCYSE CONCYSE is an interactive, structural, finite element program for nuclear power plant applications. The program is used in the analysis of structural members, baseplates, and welds. CONCYSE has extensive plotting capabilities and can combine output files into a complete analysis report for the structure. CONCYSE allows the finite element model to be created and inspected interactively which reduces the possibility of analyzing an incorrect model.

Added features are an ASME, Subsection NF Code Check on the structural members and an interactive program to analyze or design the welded connections

of the structure. Verification of CONCYSE was performed by UE&C's computer service vendor and is consistent with UE&C General Administrative Procedure 19 (GAP-0019), Rev. 2, September 16, 1985.

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Page 11 3.9(B).1.3 Experimental Stress Analysis The experimental stress analysis method is not used for seismic Category I ASME Code and non-Code mechanical components. The analytical method is used instead. 3.9(B).1.4 Considerations for Evaluation of the Faulted Condition

a. Seismic Category I Components All seismic Category I mechanical com ponents were evaluated for the faulted loading conditions. In most cases, conservative stress limits such as the stress limits for the upset plant condition were applied to the faulted condition. Actual stresses are within elastic limits. The inelastic method of analysis was not used to evaluate the design of safety-related Code or non-Code items for the faulted condition. For Code or non-Code mechanical equipment supports, elastic analysis methods were used for evalua ting faulted loading conditions, and the stress limits for mechanical supports are within the elastic range of the material used in the fabrication of the support. The rules employed for evaluation of faulted condition loading effects for supports of ASME Code mechanical equipment are in accordance with either Subsection NF of the ASME B&PV Code,Section III, or AISC criteria which provide comparable stress criteria.
b. Piping 1. Faulted Condition Under Pressure-Retaining Integrity All seismic Category I piping systems were evaluated for the faulted loading condition to provide assurance of pressure-retaining integrity.

The elastic method of analysis was used according to the rules of the ASME Code,Section III, Appendix F.

The effect of the anchor displacements and temperature conditions for the faulted condition on the piping systems was evaluated, using a static method of analysis for displacement and thermal expans ion. Specifically, the following acceptance criteria were used for the various piping systems during the faulted condition: (a) Class 1 piping systems were designed and analyzed for the load combinations and stress limits of Table 3.9(B)-8.

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Page 12 (b) Class 2 and 3 essential piping systems which are required to retain functional integrity were designed and analyzed for the load combinations and stress limits of Table 3.9(B)-9. The check of equations (9), (10) and (11) of NC/ND-3652 assured that piping deformation due to anchor displacements and thermal effect of fluid or environmental temperat ure during the faulted condition will not impair the functional integrity of the systems. 2. Faulted Condition under Structural Integrity Class 2 and 3, and nonclass, nonessential piping systems which are required to retain structural integrity were designed and analyzed for the load combinations and stress limits of Table 3.9(B)-10. The check of primary stress limits assured that structural integrity of the piping will not be impaired. The effect of environmental temperature on the piping system was evaluated in terms of additional loads on supports and

anchors, due to expansion of piping predicted on the basis of elastic

properties of the piping. The expansion stress limits were not checked for piping in this category since structural deformation induced by self-limiting load is permitted (expansion stresses which result from the constraint of free-end displacement). 3. Faulted Condition under Pipe Break Refer to Subsections 3.6(B).2 a nd 3.6(N).2 for a discussion on this subject. c. Supports For evaluation of supports under the faulted condition, refer to Subsection 3.9(B).3.4.

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Page 13 3.9(B).2 Dynamic Testing and Analysis 3.9(B).2.1 Piping Vibration, Thermal Expansion and Dynamic Effects All ASME Code Class 1, 2 and 3 piping for S eabrook Station has been designed, analyzed, fabricated and erected in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code. Design and analysis of the piping systems included consideration of thermal, deadweight, pressure, seismic and operating transient loads.

The following tests will be performed on ASME Class 1, 2 and 3 systems, high-energy systems located inside seismic Category I structures, high-energy systems whose failure could reduce the function of any seismic Category I feature to an unacceptable level, and seismic Category I portions of moderate energy systems located outside containment.

a. Vibration As part of the preoperational testing program, transient and steady-state vibration of high-energy critical piping will be monitored and evaluated. The following testing program will be implemented to test and evaluate the vibration effect:
1. A list of the systems, flow modes of operation and transients to which each system will be tested is contained in Table 3.9(B)-1. 2. Small bore piping and instrumentation lines connected to the main piping will be included in the testing program. 3. For all system lines within the scope of the systems identified in Table 3.9(B)-1, the vibration levels will be analyzed. At selected locations, the peak-to-peak displacements will be measured. Inaccessible locations and noninstrumented transients will be visually observed. 4. If measured parameters exceed the acceptance criteria, the effect of the vibration on design will be determined by further evaluation. Areas of evaluation are as follows: (a) Fluid system parameters, pressure, flow rate and temperature (b) Frequency of vibration. Following further evaluation, corrective action will be initiated until the system meets the acceptance criteria.

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Page 14 5. For steady-state vibration, the piping break stress (zero to peak) due to vibration only (neglecting pressure) will not exceed 12,000 psi for austenitic stainless steel and 7,690 psi for carbon and low alloy steels. These limits are below the piping material fatigue endurance limits as defined in the design fatigue curves in Appendix I of ASME code for 10 6 cycles. b. Thermal Expansion During hot functional testing, the temperature of the reactor coolant and main steam piping will be raised to the operating temperature in increments of 100 F. At selected temperatures, the thermal expansion of the piping will be checked at predetermined points. 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. If the support system is found to be functioning properly, but thermal expansion measurements vary from predicted values by more than a reasonable tolerance, then the analysis will be reviewed to explain the anomaly. Reasonable tolerances, for the purpose of these tests, are as follows: 1. For expected motions of 1/8 inch or less, no measurements are anticipated.

In this case a tolerance of +/- 1/8 inch is acceptable. 2. For expected motions larger than 1/8 inch but less than 2 inches, a tolerance of +/-50 percent is acceptable. 3. For expected motions greater than 2 inches, a tolerance of +/-25 percent of the expected motion is acceptable. Systems and locations requiring monitoring of thermal expansion during startup functional testing are listed in Table 3.9(B)-2.

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Page 15 c. Dynamic Effects During the preoperational test and initial startup programs, the pumps will be started and stopped and valves will be st roked to verify proper operation. The effects of the transient vibrations and shocks resulting from these operations will be visually observed to verify that no excessive motion of piping or equipment results. If excessive motions are obser ved as a result of these operations, the support systems will be adjusted to eliminate the unacceptable motion. In cases where sequence of operations or operating procedures are found to cause

unacceptable conditions, procedures will be changed to eliminate the unacceptable conditions. For operating transient vibration, the piping bending stress, (zero to peak) due to the operating transient vibration, will be co rrelated with the calculated stress level from the piping stress report to ensure that the calculated total primary stress level is maintained within acceptable code levels. Where testing results indicate unacceptable levels, action will be taken to reduce or eliminate the effect of the

operating transient vibration. The details of the piping vibration, thermal expansion and dynamic effects testing will be incorporated into the individual test procedures for conducting these tests, as abstracted in Chapter 14. Specific information and data will be generated for the preparation of piping tests. This will include the different flow modes, identification of the selected locations for visual inspections and measurements, the acceptance criteria, and the possible co rrective actions if excessive vibration occurs.

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Page 16 3.9(B).2.2 Seismic Qualification Testing of Safety-Related Mechanical Equipment Seismic Category I safety-related mechanical equipment has been designed to withstand the combined normal operating accident loads concurre nt with the effects of the Safe Shutdown Earthquake (SSE) or the Operating Basis Earthquake (OBE). Verification of the adequacy of the equipment to withstand these combined loads has been demonstrated by analysis, by testing, or a combination of analysis and testing. Selection of the method of qualification was determined by the type, size, shape and complexity of the equipment being considered. Operability of active Class 1, 2 or 3 valves, active Class 2 and 3 pumps, and their vital auxiliary equipment has been demonstrated by the combina tion of analysis and testing. Results of the qualification tests are given in Subsection 3.9(B).3.2. The seismic qualification procedures used for the operability verification of other active mechanical equipment are similar to the procedures described in Subsection 3.9(B).3.2 for pumps and valves.

For the mechanical equipment which is mechanically or structurally too complex, i.e., the seismic response cannot be adequately predicted analytically, testing procedures similar to those described in Section 3.10 were used to demonstrate the equipment operability and adequacy.

The structural and functional integrity of nonactive seismic Category I mechanical equipment has been demonstrated by one of the following two methods: a. By analytical methods, to satisfy the stress criteria applicable to that specific piece of equipment b. By seismic testing, to show that the equipment retains its structural and functional integrity under the simulated test environment This method is not applicable to pressure retaining boundaries of equipment since the stresses at the pressure boundary must be evaluated by analysis, as required by Section III of the ASME Boiler and Pressure Vessel Code. The analytical or testing qualification methods discussed above are similar to the procedure described in Subsection 3.9(B).3.2, except that no functional verification is required.

Generally, the test procedures are based upon IEEE STD 344-1975 using random input motion. For the purely mechanical equipment, test procedures for a limited number of components were based upon IEEE STD 344-1971, using sinusoidal input motion and including a 1.5 factor applied to the motion. This results in compliance with IEEE STD 344-1975.

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Page 17 3.9(B).2.3 Dynamic Response Analysis of Re actor Internals under Operational Flow Transients and Steady-State Conditions Refer to Subsection 3.9(N).2.3. 3.9(B).2.4 Preoperational Flow-Induced Vibr ation Testing of Reactor Internals Refer to Subsection 3.9(N).2.4. 3.9(B).2.5 Dynamic System Analysis of the Reactor Internals under Faulted Conditions Refer to Subsection 3.9(N).2.5. 3.9(B).2.6 Correlations of Reactor Internals Vibr ation Tests with the Analytical Results Refer to Subsection 3.9(N).2.6. 3.9(B).3 ASME Code Class 1, 2 and 3 Components, Component Supports and Core Support Structures 3.9(B).3.1 Loading Combinations, Design Transients, and Stress Limits The load combinations and the design stress limits associated with the plant operating conditions which are applied to the design and analysis of the ASME III Code-constructed items, other than the NSSS items, are defined herein. The plant conditions considered were normal operation, postulated accidents and specified seismic events.

Design transients are further discussed in Subsection 3.9(B).1.1. The requirements of ANSI/

ANS-51.1-1983 have been satisfied by use of plant conditions and allowable stress limits imposed on active and nonactive components.

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Page 18 For the non-ASME component members (other th an bolts) of the ASME III Code-constructed items, the design criteria limit the principal stresse s to 0.6 Fy for the plant upset conditions, and to 0.9 Fy for the plant faulted condition. Refer to Subsection 3.9(B).3.4c for the bolt design criteria. a. Valves, Pumps, Heat Exchangers, and Tanks The loading conditions considered (whe re applicable) for the design of ASME Class 1, 2 and 3 components included, but were not limited to, loading effects resulting from: 1. Internal and external pressure

2. Dead load, i.e., weight of the component and normal contents, including additional pressure due to static and dynamic head of liquid 3. Superimposed loads caused by other components, such as nozzle loads
4. Environmental loads, wind loads, snow loads, and seismic loads for both an OBE and a SSE 5. Valve thrust and moments
6. Thermal and thermal transients (for Class 1 components only). The loading combinations considered (where applicable) in the design and analysis of the ASME Code Class 1, 2 and 3 and certain non-Code safety-related components were categorized with respect to plant operating conditions defined as normal, upset, emergency and faulted conditions, as identified in Table 3.9(B)-3. The corresponding stress limits for each category of plant operating condition are presented in Table 3.9(B)-4 for nonactive pumps, Table 3.9(B)-5 for nonactive valves, Table 3.9(B)-6 for nonactive Class 1 valves and Table 3.9(B)-7 for ASME Code Class 2 and 3 pressure vessels and storage tanks. The stress limits for active pumps and valves are discussed in Subsection 3.9(B).3.2. The stress limits established for the various components

are sufficiently low so that violation of the pressure boundary will not occur.

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Page 19 b. Piping Systems, Including In-Line Valves The safety-related piping systems have b een designed to satisfy the appropriate stress limits of the ASME III Code and those of Regulatory Guide 1.48, as delineated below: 1. For those piping systems that constitute a portion of the reactor coolant pressure boundary and have been designated as ASME III, Class 1 lines, the load combinations and stress limits for various plant operating

conditions are presented in Table 3.9(B)-8. The following are the ASME Code Class 1 pipes qualified by the original A-E, UE&C (see Subsection 3.9(N).1.1 for design transient list applicable to Class 1 components):

Line No. Line Size Line Description

P&ID 91-1 1" Reactor vessel vent line RC-20845 91-2 1" Reactor vessel vent line RC-20845 328-6 2" From seal inject. Filters to RC-P-1A CS-20726 328-7 11/2" From seal inject. Filters to RC-P-1A CS-20726 329-4 2" From seal inject. filters to RC-P-1B CS-20726 329-5 11/2" From seal inject. filters to RC-P-1B CS-20726 330-4 2" From seal inject. filters to RC-P-1C CS-20726 330-5 11/2" From seal inject. filters to RC-P-1C CS-20726 331-4 2" From seal inject.filters to RC-P-1D CS-20726 331-5 11/2" From seal inject. filters to RC-P-1D CS-20726 80-1 6" Pressurizer vent line RC-20846 80-2 3" Pressurizer vent line RC-20846 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 14 Section 3.9(B)

Page 20 Line No. Line Size Line Description

P&ID 80-15 6" Pressurizer vent line RC-20846 80-6 3" Pressurizer vent line RC-20846 74-1 6" Suction line of pressurizer relief valve RC-20846 75-1 6" Suction line of pressurizer relief valve RC-20846 76-1 6" Suction line of pressurizer relief valve RC-20846 Westinghouse has responsibility for Class 1 component core support structures and specific Class 1 pi ping. UE&C has responsibility for pressurizer safety relief line, the reactor coolant system drain line and Class 1 reactor coolant pump seal piping. 2. For those essential piping system s which have been designated as ASME III, Class 2 and 3, and which ar e required for safe shutdown of the reactor, the load combinations and stress limits for various plant operating conditions are presented in Table 3.9(B)-9. 3. For those nonessential piping system s which have been designated as ASME III Class 2 and 3, but which are not required for safe shutdown of the reactor, the load combinations and stress limits for various plant operating conditions are presented in Table 3.9(B)-10. Definitions of the symbols and notations used in Table 3.9(B)-8, Table 3.9(B)-9 and Table 3.9(B)-10 are contained in Table 3.9(B)-11. 4. For those piping systems which are non-ASME III, design criteria were specified so that structural integrity of such systems could be maintained during the most adverse plant condition.

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Page 21 For any of the above piping systems which contained in-line components, such as valves, flow elements, strainers, etc., the loads imposed on such items by the piping were verified to be less than the limits established by the vendor. If operators were included on such valves, the seismic accelerations imposed were verified to be less than the levels to which the unit was qualified, either on a structural integrity criteria for nonactive valves or on an operability criteria for active valves. The latter criteria are discussed further in Subsection 3.9(B).3.2. Analyses of all seismic Category I piping systems have been conducted using either the ADLPIPE, ADLPIPE-2 or IMAPS computer program. In each of these programs, the mathematical models employed to represent the piping system consisted of lumped masses interconnected by beam elements whose elastic properties matched those of either the piping section, or an in-line component, such as a valve. Support elements and equipment attachment points were included in such models. Lumped masses, offset from a section centerline, were included, when necessary, to model valve operators. Structural boundaries of a mathematical model were defined by equipment connections, by support system anchors or by restraints which formed the boundary between seismic Category I and nonseismic Category I piping systems.

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Page 22 3.9(B).3.2 Pump and Valve Operability Assurance The pumps and valves identified as active, ASME Section III components that must perform a mechanical motion during the course of performing their safety function in shutting down the plant to and maintaining it in a cold shutdown or in mitigating the consequences of a postulated event, are listed in Table 3.9(B)-26 and Table 3.9(B)-27, respectively. These active pumps and valves are classified as seismic Category I, and are designed to perform their intended functions during the life of the plant under all postulated plant conditions. The operability of these active pumps and valves is assured by adherence to the design limits and supplemental stress requirements specified in NRC Regulatory Guide 1.48.

Safety-related active valves are qualified by prototype testing and analysis; safety-related active pumps are qualified by analysis and functional test.

All applicable loads, such as seismic, nozzle and operating loads are considered in the test program and the analysis. Operational tests of the originally-installed components at design basis conditions were performed during plant test start-up. All originally-designated active valves were tested by vendor at full pressure, nozzle loads, and seismic loads during the operability te sts. Full flow conditions were generally not included during these tests but a differential pre ssure was applied and was effective during the opening stage of the valves. Except for the main steam isolation valve and the feedwater isolation valve, the other valves were not tested at elevated temperatures. ANSI B16.41 provided the technical guidance for qualifying equipment by similarity analysis and prototype testing. Where components are a part of a pressurized piece of equipment within the pressure boundary, testing of the assembly was performed. Components not originally-designated as active were reviewed for loading and seismic accelerations, to ensure that they meet active component requirements. By means of these programs, the structural integrity and the ability to perform the safety-related functions are assured for the active components during the postulated plant loadings. The details of the operability assurance programs are presented below.

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Page 23 a. Pump Operability Assurance Program The overall operability verification program consists of both testing and analysis. The test programs used to establish the operability of all active seismic Category I pumps include tests by the pump manufactur er prior to installation, and by the licensee after installation in the plant. The tests performed at the pump manufacturer's facility include hydrostatic tests of pressure retaining parts in accordance with the requirements of ASME Section III, Subsection NC or ND, and pump hydraulic performance tests in accordance with Hydraulic Institute standards. After the pump is installed in the plant, the pump is subjected to

pre-operational tests, including cold hydro tests and hot functional tests. The required periodic in-service inspection and operation tests, in accordance with the IST Program or other procedures, can further demonstrate reliability of the active pump for the design life of the plant. In addition to these tests, seismic Category I active pumps are analyzed for operability when subjected to the SSE seismic condition and applicable or conservative seismic nozzle loads, to ensure the pump will not be damaged during the seismic event and that the pump will continue operating after the SSE seismic condition. Three areas of analysis are performed on the pumps: the pump body, the rotating assembly and the pump supports. In the pressure boundary pump body analysis, the stress limits for all loading combinations, including faulted, are given in Table 3.9(B)-12. Table 3.9(B)-26 lists all AE-supplied active pumps. The rotating assembly is analyzed for

operability during the faulted condition by a ssuring that: (1) the deflection of the pump impeller shaft will not exceed the clearance between the impeller and impeller casing and (2) the bearing will not be subjected to excessive loads imposed by deflection of rotating assembly and by differential movement of the coupling between the pump and pump driver shaft. The pump supports, including the base frame and anchor bolts, are an alyzed for dead weight, nozzle loads, operating loads and seismic loads. The stress limits for the supports are those of

AISC Manual of Steel Construction, and are described in Table 3.9(B)-12.

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Page 24 The lowest natural frequency of all active pumps, except the service water pumps, is demonstrated by test or analysis to be greater than 33 Hz. The service water pump is a long deep well pump with a natural frequency of 9.7. The pump has two lateral restraints which maintain lateral displacements within the limits of the available clearances. Additionally, all stresses are limited to 1.5 S, thereby assuring that the pump operability is maintained in the faulted loading conditions. Pumps having a natural frequency above 33 Hz, are considered to be rigid, and the problems with amplification between the component and structure are

avoided. To avoid damage during the faulted plan t condition, three areas of analysis are performed on the motor: the supports, the rotor assembly, and the motor stator frame. The supports, bolts, and the stator frame are analyzed for deadweight, operating loads, and seismic loads and the stress limits are those of the AISC Manual of Steel Construction. Deflection of the rotor shaft was compared to the clearance between the stator and the rotor, to ensure that a rubbing-type failure will not occur. The angular and parallel shaft deflections at the coupling were calculated and compared to the allowables for the coupling. Rotor shaft stresses and bearing loads were evaluated and compared to allowables for the faulted plant conditions. All motor stresses are limited to the region of elastic deformation of the material stress-strain relationshi p and thereby provide assurance that operability is maintained in the faulted condition. b. Valve Operability Assurance Program The operability assurance program for seismic Category I active valves of all pipe sizes is comprised of tests and analysis. This program provides assurance that these valves will perform their mechanical function in conjunction with a design basis accident during a seismic event. The active valves are subjected to several tests prior to installation; namely, a shell hydrostatic test to ASME Section III requirements, seat and disc hydrostatic tests, and functional tests. After installation, preoperational tests are performed. Periodic in-service inspections

and periodic in-service testing, in accordance with the IST Program, or other Station procedures, or Technical Specification surveillances, further verify and assure the functional ability of the safety-related active valves.

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Page 25 The valve body and other pressure retain ing parts of active valves are designed and analyzed by considering operating loads and seismic-induced nozzle loads.

For valves with extended structures, an analysis of the extended structure is performed applying static, equivalent seismic loads of 3g fo r each of the three principal axes acting at the center of gr avity of the extended structures. The maximum allowable stress limits applied in these analyses demonstrate structural integrity and compliance with the limits specified by the ASME Section III Code for the particular ASME Class of valve analyzed. Stress limits for all loading combinations are presented in Table 3.9(B)-13 for Class 2 and 3 safety-related active valves and Table 3.9(B)-14 for Class 1 safety-related active valves.

Table 3.9(B)-27 lists all AE-supplied active valves. The valve body for active and nonactive valves is qualified by analysis and accounts for the interface loading imposed by the actuators. The valve actuators for active and nonactive valves are qualified by tests in accordance with IEEE 323-1974 and IEEE 344-1971. However, a 1.5 factor is applied to the sinusoidal input motion and, therefore, compliance with IEEE 344-1975 is achieved. The response of equipment at the resonant frequency at 5 percent damping for a continuous sinusoidal input is amplified approximately ten times, compared with an amplification of three times for a random motion input. By applying a factor of 1.5 to the sinusoidal input, the comparative response is 10/3 x 1.5 = 5 to 1, which is conservative. In addition to the above functional tests and analyses, representative originally-designated active valves of each design type with overhanging structures were tested to verify operability during simulated seismic events by demonstrating operational capabilities within the specified limits.

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Page 26 Functional specifications for originally-designated active valve assemblies were not prepared, but the requirements of R.

G. 1.148 were contained within the design specifications and system specifications. The requirements of ANSI B16.41 were not part of a specific test program but a ll of the individual tests defined, except for the vibration endurance tests, were performed as part of the series of tests comprised of vendor hydro tests and seismi c tests and plant startup testing. The

valve(s) chosen for the parent valve(s) for vendor seismic testing generally complied with the size extension limitations of 200 percent to 50 percent except as follows: (1) Posi-Seal butterfly valves, Class 2 and 3, 150 lb. carbon steel body with matrix operator, sizes 14 to 36 inches, were qualified by tests performed on a 30-inch valve. (2) Walworth gate valves, Class 2 and 3, 150 lb. carbon steel body with Limitorque operator, sizes 3 to 16 inches, were qualified by tests performed on 8 and 16 inch valves. Although the 3-inch valve is below the 50 percent criteria, evaluation of the valve dimensions indicate sufficient conservatism so that operability is assured. The testing was conducted on a representative number of valves. Valves from each of the design types were tested, and the valve sizes which cover the range of sizes in-service were qualified by the tests. The test results were used to qualify all valves within the intermediate range of the installed sizes. The testing procedures and basic requirements for the operability verification of

originally-designated active valves with overhanging structures are described

below: 1. The valves are designed to have a fundamental frequency which is greater than 33 Hz. This is shown by suitable test or analysis. 2. The actuator and yoke of the valve system were statically loaded to seismic loads equivalent to 3g for each of the three principal axes. The static loads were applied at the cente r of gravity of the extended structure along the direction of the weakest axis of the yoke if the two horizontal seismic loads are combined as one static load input. The design pressure of the valve was simultaneously applied to the valve during the static

deflection tests. The design pressure of the valve and the six components of specified nozzle loads were simultaneously applied to the static

deflection tests.

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Page 27 3. The valve was then operated while in the deflected position, i.e., from the normal operating mode to the faulted operating mode. The valve had to perform its safety-related function within the specified time limits. 4. Alternatively, the operability verification of the valve, including the actuator and all other appurtenances, was demonstrated by satisfying applicable vibration test requirements. 5. Motor operators and other electrical appurtenances necessary for valve operation are qualified as operable during the seismic event by testing and/or analysis in accordance with the requirements of IEEE 323 prior to

installation on the valve. Since the accelerations used for the st atic valve qualification are 3.0g horizontal and vertical, the piping designer maintains the motor operator accelerations (or offset mass acceleration) to these levels with an adequate margin of safety. If the frequency of the valve, determined by test or analysis, is less than 33 Hz, an analysis of the valve is performed to determine the acceleration value applicable

to the extended portion of the valve. The analysis provides the amplification of

the input acceleration considering the natu ral frequency of the valve and piping along with the applicable floor response spectra. The computed accelerations at the extended portion of the valve are then compared with the 3g values used in

the static load tests or the shake table test results outlined in step 4 above to assure the operability of these active valves for the applicable seismic loads. The structural integrity and operability verification of the seismic Category I active check valves are ensured by a combination of the following tests and

analysis: 1. Stress analysis including the applicable seismic loads for valve parts that could affect the operability of the valve 2. In-shop hydrostatic test and in-shop seat leakage test

3. Periodic in situ valve inspection to assure the functional ability of the valve.

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Page 28 The seismic Category I safety relief valv es are subjected to stress analyses for parts that could affect the operability of the valve for the applicable seismic loads and operating loads. These valves are al so subjected to in-shop hydrostatic test, seat leakage test and periodic in situ valve inspection. In addition to these tests, a representative number of originally-designated active valves were qualified by shake table test. Random seismic input is applied to the valve and the pressure is increased until the valve mechanism actuate

s. Successful actuation within the design requirements of the valve is demonstrated to assure the functional capability of the valve. c. Operability Assurance Program Results for Active Pumps
1. Pumps (a) Primary Component Cooling Water Pumps The structural integrity and operability of the primary component cooling water pumps have been demonstrated by analytical method. The analysis was performed by using the ASME III 1974

Code, AISC criteria (support only), plus the equipment specification. The pumps were analyzed with a combined load consisting of deadweight, seismic loads and nozzle loads.

Ingersoll-Rand Co. Report Numb er EAS-TR-7535N shows that the 14x23-5 pump satisfies all of the applicable structural integrity and operability requirements of ASME Boiler and Pressure Vessel

Code,Section III, and Regulatory Guide 1.48. A lumped mass dynamic model was used with the ANSYS Computer Program to determine the fundamental natural frequencies. The calculated first mode natural frequency is 46.19 Hz, so the pump assembly is essentially rigid. A static analysis was performed to determine the stresses and deflections which will result from the application of the actual plant suction

and discharge nozzle loads in conjunction with internal pressure, deadweight, and seismic acceleration loads. The resulting stresses, loads and clearance requirements are summarized on

Table 3.9(B)-15.

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Page 29 To complete the operability assurance program, the Westinghouse-supplied motor was qualified for Class 1E electric equipment, as defined in IEEE 323-1971 and IEEE 344-1971. The motor was analyzed using AISC criteria with a combined load consisting of deadweight, torque and seismic loads. (b) Containment Spray Pumps An operability assurance analysis was provided for the Bingham-Willamette 6x10x14B - CD containment spray pump-motor set. The applicable codes are ASME III, 1974, AISC

criteria (support only), and the equipment specification. The pump and motor set was analyzed for combined loading consisting of deadweight, seismic loads, and nozzle loads. In addition, the

operating torque was considered in the shaft analysis. The natural frequencies of both the containment spray pump and the motor were computed, and all frequencies were found to be greater than 33 Hz; therefore, the pump assembly is considered essentially

rigid. A static analysis was used to determine the stresses and deflections which result from the application of nozzle loads, deadweight and seismic loads. A summary of the deflections and stresses is given in Table 3.9(B)-16, which shows th e stresses and deflections are well within allowable limits. The deflection of the pump shaft was compared to the minimum clearance. The motor rotor deflection was compared to the air gap. The coupling deflection and the angular misalignment were compared to the vendor-specified allowable limits. The Westinghouse-supplied motor was analyzed using AISC criteria, with a combined load consisting of deadweight, torque and seismic loads.

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Page 30 (c) Emergency Feedwater Pumps Detailed seismic analyses were performed for the two emergency feedwater pumps, one of which us es an Ingersoll-Rand Model 4x9 NH-10 turbine-driven pump and the other, a model 4x9 NH-10 motor-driven pump. The pumps were each analyzed for a combined load consisting of deadweight, seismic loads and nozzle

loads. Ingersoll-Rand Co. Report Numbers EAS-TR-8001 and EAS-TR-7925 established that both pumps were found to satisfy all of the applicable structural integrity and operability requirements of the ASME III C ode, AISC criteria (support only), Regulatory Guide 1.48 and equipment specification. A mathematical model was developed and dynamic frequency analysis was performed using the computer code ANSYS. The calculated first mode natural frequency is 48 Hz for the turbine-driven pump assembly and 49.3 Hz for the motor-driven pump set. The two pump assemblies were considered to be rigid, and static analysis was used to determine the structural responses.

The resulting critical deflections versus clearance requirements are summarized in Table 3.9(B)-17. Testing and analysis was performed to qualify the GS-2N Terry Steam Turbine which is used to drive the 4x9 NH-10 pump. The Turbine System was tested at Wyle Laboratory using random motion test. Supplementary analysis was performed for the pressure-retaining parts and other essential components of the steam turbine. The motor was analyzed using AISC criteria for a combined load consisting of deadweight, torque and seismic loads. Westinghouse Seismic Qualification included in Ingersoll-Rand Co. Report

  1. 75F32233 established that the structural integrity and operability

are satisfied.

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Page 31 (d) Service Water Pump Seismic analyses of the service water pump (Johnston Pump Company Model 30 DC, 2 stage vertical pump) were performed to verify operability and ability to withstand the postulated seismic events. A multi-degree-of-freedom lumped mass model was developed to represent the pump-column assembly and the discharge head-motor combination. The model consists of a series of mass points connected by weightless two-dimensional beams between nodal points. The shaft assembly between the motor and bowls is supported at each bearing along the column, and linear

springs were used to represent the bearing stiffness. The vertical flexibility of the base plate was represented by a linear spring, and a torsional spring was used to represent the spring rate of the base plate subjected to the applied moment. The natural frequencies, deflections, loads, and stresses were obtained with the aid of computer program ANSYS. The calculated fundamental frequency of the pump is 9.613 Hz for the entire structure. The deflection of the impeller relative to the pump casing was determined by analysis to be 0.002 in. A deflection of .002 inch translates into a natural frequency of 70 Hz. The pump is, therefore, cons idered rigid and per FSAR commitment, seismic operability testing is not required. A dynamic analysis was performed to determine the structural responses due to the seismic loads for the vertical and lateral directions. The nozzle loads and other normal operating loads were combined with the seismic loads. A supplemental analysis was performed by Ingersoll-Dresser Pump for the IDP Model 42APK service water pump, which utilizes the previously existing discharge head, motor, and baseplate (IDP Technical Report 2 000-17). The resulting stresses were compared to the allowable stresses, and the resulting deflections were compared to operating clearances or other operating criteria. A summary of the stresses and deflections is

given in Table 3.9(B)-18.

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Page 32 The seismic, stress and deflection analysis of the General Electric Company's 600 HP, Model 5K6338XC102A, pump motor was performed by McDonald Engineering Analysis Company. The

analysis was directed toward pr oving both the structural integrity and functional capability of the motor. The seismic loads, including the reduced magnetic and centrifugal loads, were imposed on the mathematical model of the motor assembly, and

the resulting stresses and deflections were compared to the code allowables and the operating clearances or other limiting criteria.

McDonald Engineering Company's Report, ME 267, established that the motor satisfies all of the applicable requirements. (e) Cooling Tower Pump The seismic analysis of the cooling tower pump was performed for the Model 33NLC Johnston vertical pumps. The analytical approach was essentially the same as for the service water pump.

Johnston Seismic Report ME-772 shows that the pump satisfies all of the applicable structural integrity and operability requirements of ASME III and the specification requirements. The deflection of the impeller relative to the pump casing was determined by analysis to be 0.0019 in. The pump is, therefore, considered rigid and per FSAR commitment, seismic operability

testing is not required. A dynamic analysis was performed to determine the stresses and deflections which will result from the application of the seismic loads, deadweight and nozzle loads. A supplemental analysis was performed by Ingersoll-Dresser Pump for the IDP Model 29LKX cooling tower service water pump, which utilizes the previously existing discharge head, motor, and baseplate (lDP Technical

Report 2000-16). The resulting stresses, loads and clearance requirements are summarized in Table 3.9(B)-19. The General Electric-supplied 800 HP Model 5-K 6339XC179A motor was analyzed by McDonald Engineering Analysis Company.

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Page 33 It is similar to the motor used for service water pump. Stress Report #266 established that the motor satisfies all of the applicable requirements. (f) Diesel Fuel Oil Transfer Pump Operability of the diesel fuel oil transfer pump under the most adverse applicable combination of normal loads, nozzle loads, and seismic loads has been analytically demonstrated by the pump manufacturer, Delaval IMO Pump Division. The pump assembly consists of a Delaval IMO screw-type pump (N3DBS-187) and a Westinghouse electric motor mounted on a common bedplate. The natural frequency of the unit was calculated; the fundamental natural frequency was determined as 228 Hz. Static analysis was used to determine the structural responses and the resulting stresses; deflections are summarized in Table 3.9(B)-20. The Westinghouse supplied 2 HP, Type T, fan-cooled AC motor was analyzed by Westinghouse Medium Motor & Gearing Division. Westinghouse Qualification Document MM-9112 and Certification letter dated February 13, 1981 assure that the motor satisfies all of the applicable requirements. The seismic analysis report is available for review at the Westinghouse Medium Motor

& Gearing Division. (g) Spent Fuel Pool Cooling Pumps A detailed seismic analysis was performed for the Model 6X8X12 CF spent fuel pool cooling pumps supplied by Bingham-Willamette. This analysis verified the equipment's

ability to resist the design basis seismic event by demonstrating the adequacy of critical components when subjected to normal and seismic loads. The applicable design codes were ASME Section III 1974 edition for pressure boundary components and AISC 1970 edition for structural support members. Loads considered in this analysis include nozzle, motor torque and seismic inertia. The natural frequencies of the pump and motor were calculated and all frequencie s were found to be significantly greater than 33 Hz, refer to Tabl e 3.9(B)-29. Therefore, the spent fuel pump/motor assembly is considered rigid.

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Page 34 The stresses and deflections for critical components of the pump/assembly resulting from the application of nozzle, deadweight, seismic inertia and motor torque loads as appropriate were determined using static methods. Critical components addressed include the shaft, pedestal weld and the pump, motor and assembly hold down bolts. Add itionally, deflections of critical rotating components were also calculated and compared to allowable clearances. For example, it was demonstrated that the motor rotor deflection was less than the available air gap between

the rotor and stator. Likewise, the shaft coupling deflections and possible misalignment was confirmed to be within acceptable limits. The pump motor manufacturer, Westinghouse, has also performed a seismic analysis demonstrating acceptability to the specified seismic requirements. d. Operability Assurance Program Results for Active Valves The results of seismic tests and analysis of originally-designated active valves was provided in the document entitled, "Public Service Company of New Hampshire, Seabrook Station Units 1 & 2, Seismic Qualification Review Team (SQRT) Equipment List," which was forwarded to Mr. Frank J. Miraglia, Chief Licensing Branch #3, Division of Licensing, under cover of PSNH's letter, dated May 27, 1982. Valves that are not ASME Section III or are not 1E that perform a mechanical motion to accomplish or support a safety function

  • are periodically tested per the IST Program, or other station proce dures, or Technical Specification surveillances. These valves (both BOP and NSSS) are listed in Table 3.9(B)-28.

Valves located in a harsh environment have been analyzed for environmental qualifications and are included in the associated station programs.

  • This includes some NNS valves, located in portions of systems that serve no safety function. Per the NRC safety evaluation for Seabrook Station (NUREG 0896), the NNS main feedwater regulating and bypass valves provide for redundant feedwater line isolation function to mitigate a main steamline or feedwater line break. NURGE 0896 also indicates that these fail-closed valves serve as an acceptable backup to the main feedwater isolation valves. In response to NRC Generic Letter 96-06, it was identified that several relief valves in NNS piping protect containment penetration isolation valves from an overpressure condition. Subsequently, it was identified that a relief valve in a normally isolated NNS piping section protect ECCS boundary isolation valves from an overpressure

condition. To ensure continue reliability of the above NNS relief valves, they ar e being periodically tested.

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Page 35 3.9(B).3.3 Design and Installation Details for Mounting of Pressure Relief Devices The installation and design of pressure relief devices comply with the rules of ASME III, Paragraph NB-7000, and NRC Regulatory Guide 1.67. a. Overpressure Protection for Reactor Coolant Pressure Boundary (RCPB)

The pressurizer in the Reactor Coolant System is provided with three safety valves and two power-operated relief valves for overpressure protection. These valves discharge through a closed piping system to the pressurizer relief tank, where the steam is condensed and cooled by mixing with water. The piping system and supports are designed to satisfy the following design criteria: 1. Stress limits for load combinations listed in Table 3.9(B)-8 for safety Class 1 piping from the pressurizer to the safety and relief valves 2. Stress limits for load combinations listed in Table 3.9(B)-9 for nonsafety class piping downstream of the safety and relief valves to the pressurizer relief tank 3. Load limits on pressurizer vessel nozzles as established by the manufacturer of the pressurizer vessel 4. Load limits on valve connections as established by the manufacturer of the valves. The three safety valves are mounted on th e pressurizer nozzles with the short inlet pipe and elbow necessary to position the valves vertically. The total length of pipe, elbow and weld-neck flange is approximately 24 inches and is as short as possible to minimize the pressure drop on the inlet side of the valve. When the valves open, the dynamic effects from the flow of water and steam are included in the design analysis. These transient load effects on the piping system, upstream and downstream of the safety and relief valves, have been evaluated in the following manner: 1. Safety Valve Piping System A static analysis was performed for the Safety Valve Piping System in which the peak transient loads obtained from a RELAP 5 analysis and multiplied by a dynamic load factor (DLF) were applied. The Pressurizer Safety Valve Piping System contains no water seals nor is subjected to water slugs.

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Page 36 2. Pressurizer Relief Valve Piping System Both static and time history analyses were performed for the Pressurizer Relief Valve Piping System using transient loads obtained from a RELAP 5 analysis. The Pressurizer Relief Valve Piping System contains water seals and is subjected to water slugs. The effects of these two items were fully accounted for in the RELAP 5 analysis. The RELAP 5 computer code, (Reference 1), was used to generate thermal hydraulic characteristics of the flow along the piping system, from which tables of the wave force versus time for each leg have been derived. To evaluate piping stresses and support loads, the maximum force for each leg has been selected and applied statically to the piping system in the most conservative fashion using a dynamic load factor (DLF) based on the valve opening time and the system dynamic characteristics. In cases where time history analyses were performed, the appropriate thermal hydraulic forcing function was applied to the applicable pipe segment.

The developed stresses and loads on nozzles were combined with the other applicable loads from Table 3.9(B)-8 and Table 3.9(B)-9. These were compared with the allowable stresse s and allowable nozzle loads. The simultaneous discharge from all valves has been assumed in the thrust analyses. b. Overpressure Protection for the Secondary (Main Steam) System A multiple-valve installation, comprised of five safety valves, is provided in each of the four main steam lines. The valves are installed on main steam piping headers, outside of the Containment Building in a piping chase between the containment penetration and the main steam isolation valves. The safety valve discharge side is configured to minimize reaction forces at the valve branch/main header intersection point. The vertical branch line from the main steam piping

header to each individual valve has a forg ed flange and sweepolet welded to the header. Safety valves are bolted directly to the flanges. The effect of the valve discharge transien t was obtained by static application of an assumed discharge force, as obtained from the valve manufacturer, with a dynamic load factor DLF based on the system dynamic characteristics. It has been assumed that all five valves discharge simultaneously. The system of piping supports and rigid restraints limits both dynamic and static loadings to the piping system to code allowable stresses for the load combinations listed in Table 3.9(B)-9.

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Page 37 c. Safety and Relief Valves for Various Auxiliary Systems Mounting of safety and relief valves on auxiliary piping systems uses standard piping components: flanges, buttwelded or socket-welded tees, weldolets and sockolets for pipe branches to the va lves. The valves and valve discharge piping utilize flanged joints, buttwelded a nd socketwelded connections. Branch connections are qualified using code sta ndard calculations for tees with proper intensification factor (ASME III, Table NB-3682.21 or NC-3652-4). The alternative method for branch qualification is the Bijlaard method using the SPHNOZ/CYLNOZ computer program. The load combination for calculating stresses is according to Table 3.9(B)-9. These were compared with the allowable stresses. The following basic installation of safety valves outlet piping has been utilized: 1. Open discharge with the minimum pi ping length, or no piping attached to the valve discharge and discharging to atmosphere. 2. Open discharge system, discharging directly to atmosphere through individual piping systems, or common header combining discharge from

several valves. 3. Closed discharge system, dischargi ng to a container through an individual piping system, or common header combining discharge from several valves. The discharge reaction forces have been obtained from one of the following sources: 1. Valve manufacturer

2. For open steam discharge, from nonmandatory Appendix "O" of ASME III 3. For open and closed discharge systems with piping system connected to lve discharge, from UE&C proprietary computer programs MADIS, VALCLO, ELBFOR, and WATER (described in Subsection 3.9(B).1.2).

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Page 38 The stress analyses of piping systems downstream and upstream of a valve have been obtained by applying the reaction forces statically with a dynamic load factor (DLF) as appropriate based on the system dynamic characteristics. Piping stresses from the safety valve discharge have been combined with the stresses from other applicable loads in accordance with the load combination from Table 3.9(B)-9. These were compared with the allowable stresses. d. Results of Analysis The results of a dynamic analysis of the various piping systems have been summarized in the tabular format represented by Table 3.9(B)7-21 and Table 3.9(B)-22. 3.9(B).3.4 Component Supports

a. ASME Code Class 1, 2 and 3 Piping Supports
1. Jurisdictional Boundaries The jurisdictional boundaries of s upports designed and fabricated to Subsection NF requirements as shown in NF-1000 for plates, welding and bolting is as follows: (a) Plates (1) Support plates that are embedded in concrete with integral embedded anchors (studs) do not fall within NF jurisdiction, whether or not they extend beyond the surface

of the concrete. (2) Loose or adjustable base plates which support only compressive loads do not fall within NF jurisdiction. (3) Loose plates that are welded to component supports, such as surface-mounted plates, fall within NF jurisdiction. (b) Welding Welds used to attach NF supports to building steel, supplementary steel, or intervening members fall within NF jurisdiction.

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Page 39 (c) Bolting (1) Embedded custom-designed anchor bolts are designed and fabricated to AISC requirements and also to the additional materials, certification and NDE requirements of

Subsection NF. (2) Expansion anchors which are manufactured and stocked as catalogue items do not fall within NF jurisdiction.

2. Loads The loads (and movements of components) considered in the analysis of ASME Code Class 1, 2 and 3 piping supports are shown in Table 3.9(B)-23. The loads were combined to determine a worst case in each direction for normal and upset condition and faulted condition.

Friction forces due to thermal movements of the piping bearing on the support structures were considered by calculating a friction force using a coefficient of friction of 0.35, or applying the thermal displacement to the support structure in those cases where the support structure is flexible. 3. Types of Supports, Design and Service Limits (a) Linear Type Supports Stresses were calculated using linear elastic analysis in accordance with ASME Section III, Division 1, Subsection NF. The maximum allowable stresses used were: (1) For normal and upset load conditions, the allowable stresses were in accordance with ASME,Section III, Division 1, Subsection NF, and Appendix XIII. (2) For the faulted load condition, the allowable stresses used were in accordance with the ASME Code, Appendix F and USNRC Regulatory Guide 1.124 for essential systems.

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Page 40 (3) The stiffness values used for support design were:

Pipe size (OD, in.) Stiffness (lb./in.)

Up to 21/2 1x10 4 21/2 to 6 1x10 5 Above 6 1x10 6 In those cases where the support stiffness was less than that specified above, the piping analysis was reviewed to determine the impact on the component. (4) Component supports are designe d to be in the rigid range (natural frequency fn33 Hz). In cases where the frequency is less than 33 Hz, the analysis of the piping system was reviewed to assure that the piping analysis remained valid. (5) The thermal movement of the component at the support was accommodated through clearance included in the component support design. (6) Component supports are connected to concrete walls and slabs by either welding to embedded plates, or by bolting to

the concrete with either concrete expansion anchors (wedge type) or concrete inserts. The response to the NRC's IE Bulletin No. 79-02, (Reference 2), was used as a guide for

the design of the concrete expansion anchors. The maximum allowable design loads for the concrete

expansion anchors for ASME Class 1, 2, and 3 supports were developed using the manufacturer's ultimate loads and

a safety factor of 4 for worst case loading (normal and upset or faulted loads). Baseplate flexibility and shear-tension interaction were accounted for in the design of the concrete expansion

anchors.

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Page 41 (7) Lug attachments welded to Class 2 and 3 pipes are qualified by a procedure whose methodology is equivalent to, but more conservative than, that presented in Code

Case N-318-1. Local stress levels in the pipe resulting from applied lug loads are obtained by multiplying the nominal stress in the

lug at the lug/pipe interface by the appropriate B or C index (as defined in Code Case N-318-1) for each individual loading condition. The local stresses are superimposed upon the general pipe stress as determined from program

ADLPIPE to establish the total stress level in the pipe for that loading condition. Loading conditions required to be considered for Plant Normal, Plant Upset, Plant Emergency, and Plant Faulted

Operating Condition are defined (per appropriate Updated

FSAR section), and total stress in the pipe is obtained from summing the stresses for each individual loading condition that must be considered. Local stress levels determined using B indices are added to the general stress levels from ADLPIPE and this sum is compared against allowable limits to demonstrate structural integrity. For the pipe wall, local stress levels determined using C indices are added to the general stress levels from ADLPIPE, and this sum is compared against the allowable

range of stress (S h+S a). (b) Component Standard Supports Component standard supports were selected on the basis of: (1) Maximum design loads (and movements) as established in Table 3.9(B)-23 (2) Allowable loads (and movements) provided by manufacturers' component standard support catalog load capacity data sheets The allowable loads were certified by the manufacturer to be generated in accordance with ASME Section III, Subsection NF, for all load conditions.

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Page 42 4. ASME Code Class 2 and 3 Support Inspection All Class 2 and 3 supports will receive surveillance and audit inspection by the Authorized Nuclear Inspection Agency (ANIA) in lieu of in-line inspection hold points by the Authorized Nuclear Inspector (ANI). This modified program will be applied to all ASME Class 2 and 3 pipe supports and I&C supports of the following systems: AS - Auxiliary Steam CAH - Containment Air Handling CAP - Containment Air Purge CBS - Containment Building Spray CC - Component Cooling Water Primary CGC - Combustible Gas Control CO - Condensate COP - Containment Online Purge CS - Chemical and Volume Control DG - Diesel Generator DM - Demineralized Water FP - Fire Protection FW - Feedwater LD - Leak Detection MS - Main Steam MSD - Main Steam Drains S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 14 Section 3.9(B)

Page 43 NG - Nitrogen Gas RC - Reactor Coolant RH - Residual Heat Removal RMW - Reactor Makeup Water RS - Resin Sluicing SS - Sample System SA - Service Air SB - Steam Generator Blowdown SF - Spent Fuel Pool Cooling SI - Safety Injection (Emergency Core Cooling)

SW - Service Water VG - Vents WG - Waste Processing - Gaseous WLD - Waste Processing - Liquid Drains S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 14 Section 3.9(B)

Page 44 In addition, all Class 2 and 3 supports included in the following systems will receive additional surveillance insp ection (i.e., increased frequency of surveillance inspection): MS - Main Steam FW - Feedwater CS - Chemical and Volume Control SB - Steam Generator Blowdown SW - Service Water SI - Emergency Core Cooling RH - Residual Heat Removal 5. ASME Code Class 2 and 3 Support Documentation Code Symbol Stamping or executi ng NF-2 Data Reports will not be required for all Class 2 and 3 pipe supports and I&C supports. Instead, a statement will be attached to N-5 Data Reports indicating that design, procurement, fabrication, installation and examination have been performed in accordance with ASME III, Subsection NF, and confirmed by ANI surveillance and audit inspection. b. ASME Code Class 1, 2 and 3 Valve Supports ASME Code Class 1, 2 and 3 valves, including active valves, are generally supported by the pipe attached. Exceptions to this method of support occur where the valve has an excessively heavy actuator relative to the valve body weight, and/or where the pipe line size is too small to provide adequate support, and where flexible (less than 33 Hz) pneum atic operators are used. For these exceptions, the analysis and design of the supports are in accordance with the

preceding subsection on piping supports.

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Page 45 c. Mechanical Equipment Supports ASME Code Class 2 and 3 component supports, including supports for pumps, vessels, tanks and heat exchangers, are designed to withstand the loads resulting from the loading criteria and operati ng condition categories as defined in Subsection 3.9(B).3.4. The allowable stresse s defined in AISC "Manual of Steel Construction" were used for plant conditi ons associated with OBE. For plant conditions associated with SSE, the stresses were limited to 90 percent of yield stress for the material involved, or in accordance with Subsection NF of the ASME Boiler and Pressure Vessel Code,Section III. The requirements for adequacy of active Class 2 and 3 pump s upports require that both stress analysis and an evaluation of pump/motor support misalignment be performed. The supports of certain mechanical equipment purchased circa 1974 were designed in accordance with the requireme nts defined in the AISC Manual of Steel Construction. In addition, the fo llowing criteria were included in the support designs: 1. Material properties used in conjunction with the support design were obtained from the tables for material strength values in the ASME III, Subsection NA, Appendix I. 2. The allowable bolt stresses were derived from the AISC Specification, without use of one-third increase factor for Normal and Upset Conditions.

For the faulted condition the AISC allowable of 0.6 Fy was multiplied by the strength factors noted in SRP 3.8.3 and 3.8.4. 3. The loading considered in the design of the supports and anchor bolts are the same as those imposed on the components. More specifically, the appropriate loads are applied to the components and the resulting reactions are used to design the supports. 4. For the faulted condition, tensile and bending stresses were limited to 90 percent of the material yield strength and shear stresses were limited to 60 percent of the material yield strength which compare favorably with the limits defined by ASME III, Subsection NF, for faulted conditions. 5. Buckling evaluations were performed in accordance with the AISC criteria without use of increase factor for faulted conditions. 6. The highest value of KL/R is less than 20 for all mechanical components (excluding piping systems).

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Page 46 The following tabulation shows the stress limits used for various bolt materials: 1. Stress Limits for Anchor Bolts for Equipment Bolt Material Allowable Tensile Stress Allowable Shear Stress ASTM A193 Grade B7 Under 2-1/2" Fy = 105 ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 125 ksi

= 62.5 ksi

= 42 ksi ASTM A540 Grade B23 Class 4 Up to 3" Fy = 120 ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 135 ksi

= 67.5 ksi

= 48 ksi ASTM A354 Grade BD For 1/4" to 2-1/2" Fy = 130 ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 150 ksi

= 75 ksi

= 52 ksi 2. High strength bolts for equipment on st ructural steel and for steel-to-steel connections.

ASTM A325 1/2" to 1" 1-1/8" to 1-1/2" Fy = 92 ksi Fy = 81 ksi Fu = 120 ksi Fu = 105 ksi ASTM A490 1/2" to 1-1/2" Fy = 130 ksi Fu = 150 ksi S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 14 Section 3.9(B)

Page 47 All allowable tension and shear values are in accordance with Manual of Steel Construction AISC. For the faulted condition, the strength factors of 1.6 or 1.7 as noted in SRP 3.8.3 and 3.8.4 were applied to the above.

d. Snubbers
1. Design Criteria Shock suppressors, mechanical and hydraulic, used in ASME Code Class 1, 2 and 3 piping or equipment conform to the requirements of the ASME B & PV Code Section III, Subsection NF. The snubbers' intended use is as shock arrestors only, and meet the following guidelines: (a) Design life of forty years (b) Operate under environmental conditions described in the Service Environment Chart, Updated FSAR Figure 3.11-1 (c) Applicable codes are as follows: (1) Mechanical Shock Suppressors - Summer Addenda of the 1974 B & PV Code. (2) Hydraulic Shock Suppressors - 1977 Edition, Winter, 1979 Addenda. (d) DELETED (e) DELETED (f) Both mechanical and hydraulic units are designed in such a way that they will not form a rigid restraint. (g) DELETED (h) Hydraulic units are designed with a means for monitoring fluid levels in the reservoirs. (i) DELETED S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 14 Section 3.9(B)

Page 48 (j) Materials used for the fabrication of ASME Code parts shall conform to the applicable requirements of Article NF-2000 of the ASME B & PV Code. (k) Fabrication of shock suppressors shall be in accordance with the applicable requirements of Article NF-4000 of the ASME B & PV Code. (l) DELETED (m) Load ratings shall be verified in accordance with the requirements of Article NF-3260 of the ASME B & PV Code. (n) Each hydraulic snubber shall be tested in compression and tension to 10 percent of its rated load and checked for leakage of the hydraulic fluid. If fluid forms dropl ets, drips or runs off the piston

rod, the shock suppressor shall be rejected. (o) Shock suppressors' packaging shall be designed to protect against salt spray, rain, dust, water vapor, shock and vibration during shipping, handling and storage. Where possible, shock suppressors shall be packaged fully assembled in a single shipping container. (p) Mechanical units shall be designed to operate normally between 50 F and 300 F. Hydraulic units shall be designed to operate normally between 30 F and 140F with temperature excursions up to a maximum of 300 F. 2. Snubber Installation and Operability Verification (a) Pre-Service Examination A pre-service examination will be made on all snubbers. This examination will be conducted during and after snubber installation and will, as a minimum, verify the following: (1) There are no visible signs of damage or impaired operability as a result of storage, handling, or installation.

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Page 49 (2) The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.), are according to design drawings and specifications. (3) Small snubbers are not seized, frozen, or jammed by manual exercising during installation. Large snubbers (those that cannot be manually exercised) will be identified and examined for proper movement during preoperational

testing as discussed in Subsection 3.9(B).3.4d.2(b) below. (4) Adequate swing clearance is provided to allow snubber movement. (5) If applicable, fluid is to the recommended level and is not leaking from the snubber system. (6) Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire and

cotter pins are installed correctly. Prior to the performance of the thermal expansion test, an inspection of all listed snubbers covering items (1), (4) and (5) will be performed as a test prerequis ite. Snubbers which are installed incorrectly or otherwise fail to meet the above requirements will be repaired or replaced and re-exami ned in accordance with the above criteria. (b) Pre-Operational Testing During thermal expansion testing, snubber thermal movements for systems whose operating temperature exceeds 250 F will be verified as follows: (1) During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement. (2) For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.

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Page 50 (3) Verify the snubber swing clearance at specified heatup and cooldown intervals. Any discrepancies or inconsistencies shall be evaluated for cause and corrected prior to

proceeding to the next specified interval. 3.9(B).4 Control Rod Drive Systems Refer to Subsection 3.9(N).4. 3.9(B).5 Reactor Pressure Vessel Internals Refer to Subsection 3.9(N).5. 3.9(B).6 In-Service Testing of Pumps and Valves An ongoing in-service test program will be provi ded to assure the operational readiness of certain Safety Class 1, 2 and 3 pumps and valves which perform a specific function in shutting down a reactor to a safe shutdown condition or in mitigating the consequences of an accident. The in-service test program is based on the requirements given in the ASME OM Code, 2004 Edition and the requirements of 10 CFR 50.55a(f) except where specific written relief has been granted by the commission pursuant to 10 CFR Part 50, Section 50.55a(f)(6)(i). Applicability of future Code addenda will be as stated in 10 CFR 50.55a(f).

3.9(B).6.1 In-Service Testing of Pumps In-service tests, analysis and record keeping will be performed for certain Code Class 1, 2 and 3 pumps in accordance with Subsection ISTB of the Code to assess pump operational readiness and to detect changes in pump hydraulic and mechanical performance relative to reference parameters. Reference values were established during pre-service testing and will be established after major maintenance or replacement. Methods of measurement will be in accordan ce with ISTB-3500. Installed or portable instruments employed for measuring or observing test quantities will have accuracies equal to or better than that specified in Table ISTB-3510-1.

In-service test records to include test plans, documentation and required corrective action will be maintained in accordance with ISTB-9000.

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Page 51 A listing of Class 1, 2 and 3 pumps subject to in-service testing is provided in Station procedures. The Station procedures also specify the minimum test frequency, during plant operation, in which test quantities are to be measured, analyzed and documented. Plant personnel shall maintain test plans that include the type of hydraulic circuit normally used for testing. 3.9(B).6.2 In-Service Testing of Valves In-service tests, analyses and record keeping will be performed for certain Code Class 1, 2 and 3 valves in accordance with Subsection ISTC of th e Code to assess valve operational readiness. The in-service testing program for valves is detailed in Station procedures. Each valve to be tested is identified by system, valve number, code class, type, function, category and applicable tests and test frequencies.

Each valve, prior to service, was tested as required by those tests defined for each valve in the In-service Test Program, in effect at that tim

e. These pre-service tests were conducted under conditions similar to those to be experienced during subsequent in-service tests, to the maximum extent practicable. When a valve or its control system has been replaced or repaired or has undergone maintenance that could affect its performance, and prior to the time it is returned to service, it will be tested to demonstrate that the performance parameters which could be affected by the replacement, repair, or maintenance are within acceptable limits. Valves with remote position indicators, will be visually observed at least once every 2 years to verify that remote valve indications accurately reflect valve operation.

Valves which act as an isolation boundary between high pressure reactor coolant piping and adjacent low pressure systems, and whose undetected failure or degradation could lead to an inter-system LOCA, will be considered Category A or A-C valves and tested in accordance with this section and the Technical Specifications.

Records and reports of in-service valve tests w ill be kept in accordance with ISTC-9000 of the Code.

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Page 52 3.9(B).7 References

1. WREM-Water Reactor Evaluation Model, Rev. 1, NUREG-75/056, May 1975, NRC, Div. of Technical Review. 2. Public Service Company of New Hamp shire letter, dated Jan. 4, 1980, to NRC, Region I, Office of Inspection and Enforcement (response to IE Bulletin No. 79-02, Rev. 2).

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Page 1 3.9(N) MECHANICAL SYSTEMS AND COMPONENTS 3.9(N).1 Special Topics for Mechanical Components 3.9(N).1.1 Design Transients The following five operating conditions, as defined in Section III of the ASME B&PV Code, are considered in the design of the Reactor Coolant System (RCS) and RCS component supports: a. Normal Conditions Any condition in the course of start up, operation in the design power range, hot standby and system shutdown, other than upset, emergency, faulted or testing conditions. b. Upset Conditions (Incidents of Moderate Frequency)

Any deviations from normal conditions anticipated to occur often enough that the design should include a capability to with stand the conditions without operational impairment. The upset conditions include those transients which result from any single operator error or control malfunc tion, transients caused by a fault in a system component requiring its isolation from the system, and transients due to loss of load or power. Upset conditions include any abnormal incidents not resulting in a forced outage and also forced outages for which the corrective action does not include any repair of mechanical damage. The estimated duration of an upset condition shall be included in the design specifications. c. Emergency Conditions (Infrequent Incidents)

Those deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence, but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system. The total number of postu lated occurrences for such events shall not cause more than twenty-f ive stress cycles having an S a value greater than that for 10 6 cycles from the applicable fatigue design curves of the ASME Code Section III.

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Page 2 d. Faulted Conditions (Limiting Faults)

Those combinations of conditions associated with extremely low probability, postulated events whose consequences are such that the integrity and operability of the Nuclear Energy System may be impair ed to the extent that considerations of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities. e. Testing Conditions Testing conditions are those pressure overload tests including hydrostatic tests and pneumatic tests. Other types of tests shall be classified under Normal, Upset, Emergency or Faulted Conditions.

To provide the necessary high degree of integrity for the equipment in the RCS, the transient conditions selected for equipment fatigue evaluation are based upon a conservative estimate of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in a plant. To a large extent , the specific transient operating conditions to be considered for equipment fatigue analyses are based upon engineering judgment and experience.

The transients selected are representative of operating conditions which prudently should be considered to occur during plant operation, and which are sufficiently severe or frequent to be of possible significance to component cyclic behavior. The transients selected may be regarded as a conservative representation of transients which, used as a basis for component fatigue evaluation, provide confidence that the component is appropria te for its application over the design life of the plant. The following design conditions are given in the equipment specifications for RCS components.

The design transients and the number of cycles of each that are normally used for fatigue evaluations are shown in Table 3.9(N)-1. In accordance with ASME III, Emergency and Faulted Conditions are not included in fatigue evaluations. a. Normal Conditions The following primary system transients are considered normal conditions: Heatup and Cooldown at 100 F per hour Unit Loading and Unloading at 5 Percent of Full Power per Minute Step Load Increase and Decrease of 10 Percent of Full Power S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 3 Large Step Load Decrease with Steam Dump Steady-State Fluctuations Initial Random Feedwater Cycling at Hot Shutdown Loop Out of Service Feedwater Heaters Out of Service a. One Heater Out of Service b. One Bank of Heaters Out of Service Unit Loading and Unloading Between 0 and 15 Percent of Full Power Boron Concentration Equalization

Refueling Reduced Temperature Return to Power Reactor Coolant Pumps Startup and Shutdown Turbine Roll Test Primary Side Leak Test Secondary Side Leak Test Tube Leakage Test. 1. Heatup and Cooldown at 100 F Per Hour The design heatup and cooldown cases are conservatively represented by continuous operations performed at a uniform temperature rate of 100 F per hour. (These operations can take place at lower rates approaching the minimum of 0 F per hour.)

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Page 4 For these cases, the heatup occurs from ambient (assumed to be 120 F) to the no-load temperature and pressure condition and the cooldown represents the reverse situation. In actual practice, the rate of temperature change of 100 F per hour will not be attained because of other limitations such as: (a) Material ductility considerations which establish maximum permissible temperature rates of change, as a function of plant pressure and temperature, may be below the design rate of 100 F per hour. (b) Slower initial heatup rates when using pump energy only.

(c) Interruptions in the heatup and cooldown cycles due to such factors as drawing a pressurizer steam bubble, rod withdrawal, sampling, water chemistry and gas adjustments. The number of such complete heatup and cooldown operations is specified as 200 each, which corresponds to five such occurrences per year for the

40-year plant design life. 2. Unit Loading and Unloading at 5 Percent of Full Power per Minute The unit loading and unloading cases are conservatively represented by a continuous and uniform ramp power change of 5 percent per minute between 15 percent load and full load. This load swing is the maximum possible consistent with operation under automatic reactor control. The reactor temperature will vary with load as prescribed by the Reactor Control System. The number of lo ading and unloading operations is defined as 13,200. One loading operation per day yields 14,600 such operations during the 40-year design life of the plant. By assuming a 90 percent availability factor, this number is reduced to 13,200.

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Page 5 3. Step Load Increase and Decrease of 10 Percent of Full Power The +/-10 percent step change in load demand is a transient which is assumed to be a change in turbine control valve opening due to disturbances in the electrical network into which the plant output is tied. The Reactor Control System is designed to restore plant equilibrium without reactor trip following a +/-10 pe rcent step change in turbine load demand initiated from nuclear plant equilibrium conditions in the range between 15 percent and 100 percent full load, the power range for automatic reactor control. In effect, during load change conditions, the Reactor Control System attempts to match turbine and reactor outputs in such a manner that peak reactor coolant temperature is minimized and reactor coolant temperature is restored to its programmed setpoint at a sufficiently slow rate to prevent excessive pressurizer pressure decrease. Following a step decrease in turbine load, the secondary side steam pressure and temperature initially increase, since the decrease in nuclear power lags behind the step decrease in turbine load. During the same increment of time, the RCS average temperature and pressurizer pressure also initially increase. Because of the power mismatch between the turbine and reactor and the increase in reactor coolant temperature, the control system automatically inserts the control rods to reduce core power.

With the load decrease, the reactor coolant temperature will ultimately be reduced from its peak value to a value below its initial equilibrium value

at the inception of the transient. The reactor coolant average temperature setpoint change is made as a function of turbine-generator load, as determined by first stage turbine pressure measurement. The pressurizer pressure will also decrease from its peak pressure value and follow the reactor coolant decreasing temperature trend. At some point during the

decreasing pressure transient, the satu rated water in the pressurizer begins to flash, which reduces the rate of pr essure decrease. Subsequently, the pressurizer heaters come on to restore the plant pressure to its normal

value.

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Page 6 Following a step increase in turbine load, the reverse situation occurs, i.e., the secondary side steam pressure and temperature initially decrease and the reactor coolant average temperature and pressure initially decrease.

The control system automatically wit hdraws the control rods to increase core power. The decreasing pressure tr ansient is reversed by actuation of the pressurizer heaters and, eventually, the system pressure is restored to its normal value. The reactor coolant average temperature will be raised to a value above its initial equilibrium value at the beginning of the transient. The number of each operation is specified at 2000 times or 50 per year for the 40-year plant design life. 4. Large Step Load Decrease with Steam Dump This transient applies to a step decrease in turbine load from full power, of such magnitude that the resultant rapi d increase in reactor coolant average temperature and secondary side steam pressure and temperature will automatically initiate a secondary side steam dump that will prevent both reactor trip and lifting of steam generator and pressurizer safety valves.

Thus, since this plant is designed to accept a step decrease of 50 percent from full power, the Steam Dump System provides the heat sink to accept

the difference in allowable unloading rates between the turbine and the RCS. The number of occurrences of this transient is specified at 200 times or 5 per year for the 40-year plant design life. 5. Steady-State Fluctuations It is assumed that the reactor coolant temperature and pressure at any point in the system vary around the nominal (steady-state) values. For design purposes two cases are considered: (a) Initial Fluctuations - These are due to control rod cycling during the first 20 full power months of reactor operation. Temperature is assumed to vary +/-3F and pressure by +/-25 psi, once during each 2 minute period. The total number of occurrences is limited to 1.5x10 5. These fluctuations are assumed to occur consecutively, and not simultaneously with the random fluctuations.

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Page 7 (b) Random Fluctuations - Temperature is assumed to vary by +/-0.5 F and pressure by +/-6 psi, once every 6 minutes. With a 6 minute period, the total number of occurre nces during the plant design life does not exceed 3.0x10

6. 6. Feedwater Cycling at Hot Shutdown These transients can occur when the plant is at "no load" conditions, during which intermittent feeding of 32F feedwater into the steam generators is assumed. Due to fluctuations arising from this mode of operation, the reactor coolant average temperature d ecreases to a lower value and then immediately begins to return to normal no-load temperature. This transient is assumed to occur 2000 times over the life of the plant. 7. Loop Out of Service The plant may be operated at a reduced power level with a single loop out of service for limited periods of time. This is accomplished by reducing power level and tripping a single reactor coolant pump. It is assumed that this transient occurs twice per year or 80 times in the life of the plant. Conservatively, it is assumed that all 80 occurrences can occur in the same loop. In other words, it must be assumed that the whole RCS is subjected to 80 transients, while each loop is also subjected to 80 inactive loop transients. When an inactive loop is brought back into service, the power level is reduced to approximately 10 percent and the pump is started. It is assumed that an inactive loop is inadvertently started up at maximum allowable power level 10 times over the lif e of the plant. (This transient is covered under Upset Conditions.) Thus, the normal startup of an inactive loop is assumed to occur 70 times during the life of the plant. 8. Feedwater Heaters Out of Service These transients occur when one or more feedwater heaters are taken out of service. During the period of time that the heaters are out of service, it is desirable to have the operator maintain the plant at full rated thermal load. To accomplish this, the operator will:

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 8 1. Calculate the appropriate steam flow reduction, which will maintain the plant at full rated thermal load after the heater has been taken out of service.

2. Reduce steam flow by the appropriate amount (calculated in 1 above) and allow plant conditions to reach a new steady-state (approximately in 10 minutes.)
3. Take heater (or heaters) out of service. The number of occurrences of each of these transients is specified at 120 occurrences over the 40-year plant design life. 9. Unit Loading and Unloading Between 0 and 15 Percent of Full Power The unit loading and unloading cases between 0 and 15 percent power are represented by continuous and uniform ramp power changes, requiring 30 minutes for loading and 5 minutes for unloading. During loading, reactor coolant temperatures are increased from the no-load value to the normal load program temperatures at the 15 percent power level. The reverse temperature change occurs during unloading. Prior to loading, it is assumed that the plant is at hot shutdown conditions, with 32 F feedwater cycling. During th e two-hour period following the beginning of loading, the feedwater temperature increases from 32 F to 300F due to steam dump and turbine startup heat input to the feedwater.

Subsequent to unloading, feedwater heating is terminated, steam dump is reduced to residual heat removal requirements, and feedwater temperature decays from 300 F to 32 F. The number of these loading and unloading transients is assumed to be 500 each during the 40-year plant design lif e, which is equivalent to about one occurrence per month.

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Page 9 10. Boron Concentration Equalization Following any large change in boron concentration in the RCS, spray is initiated in order to equalize concentration between the loops and the pressurizer. This can be done by manually operating the pressurizer

backup heater, thus causing a pressure increase, which will initiate spray at a compensated pressurizer pressure of approximately 2275 psia. The proportional sprays return the pressure to 2250 psia, and maintain this pressure by matching the heat input from the backup heater until the concentration is equalized. For design purposes, it is assumed that this operation is performed once after each load change in the design load follow cycle. With two load changes per day, and a 90 percent plant availability factor over the 40-year design life, the total number of

occurrences is 26,400.

11. Refueling At the end of plant cooldown, the temp erature of the fluid in the RCS is less than 125F. At this time, the vessel head is removed and the refueling canal is filled. This is done by pumping water from the refueling water storage tank, which is outside and conservatively assumed to be at 32 F, into the loops by means of the low head safety injection pumps. The refueling water flows directly into the reactor vessel via the accumulator connections and cold legs. This operation is assumed to occur twice per year, or 80 times over the life of the plant. 12. Reduced Temperature Return to Power The reduced temperature return to power operation is designed to improve the spinning reserve capabilities of the plant during load-follow operations without part length rods. The transient will normally begin at the ebb (50 percent) of a load-follow cycl e and will proceed at a rapid positive rate (typically 5 percent per minute) until the abilities of the control rods and the coolant temperature reduction (negative moderator coefficient) to supply reactivity are exhausted. At th at point, further power increases are limited to approximately one percent per minute by the ability of the boron system to dilute the reactor coolant. The reduction in primary coolant temperature is limited by the protection system to about 20 F below the programmed value.

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Page 10 The reduced temperature return to power operation is not intended for daily use. It is designed to s upply additional plant capabilities when required because of network fault or upset condition. Hence this mode of operation is not expected to be used more than once a week in practice (2000 times in 40 years). 13. Reactor Coolant Pumps (RCP) Startup and Shutdown The reactor coolant pumps are started and stopped during routine operations such as RCS venting, plant heatup and cooldown, and in connection with recovery from certain transients such as Loop Out of Service and Loss of Power. Other (undefined) circumstances may also require pump starting and stopping. Of the spectrum of RCS pressure and temperature conditions under which these operations may occur, three conditions have been selected for

defining transients:

Cold condition - 70 F and 400 psig Pump restart condition - 100 F and 400 psig Hot condition - 557 F and 2235 psig For RCP starting and stopping operations, it is assumed that variations in RCS primary side temperature and in pressurizer pressure and temperature are negligible, and that the steam generator secondary side is completely unaffected. The only significant variables are the primary system flow and the pressure changes resulting from the pump operations.

The following cases were considered:

Case 1 - First Pump Startup (Last Pump Shutdown)

Variations in reactor coolant loop flow accompany startup of the first pump, both in the loop containing the pum p being started and in the other loops (loops in which the pumps remain idle). This case involves a higher dynamic pressure loss in the loop containing the pump being started, but the magnitude of the flow change is less than in Case 2. For the last pump shutdown case, the transient is the reverse of the first pump startup

transient.

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Page 11 Case 2 - Last Pump Startup (First Pump Shutdown)

This case conservatively represents the variations in reactor coolant loop flow accompanying startup of the second and third pumps. Initially, flow exists through these loops in the reverse direction as the result of starting the first pumps. The remaining pumps are then started in sequence, and a new equilibrium flow is established. The magnitude of flow reversal is

the largest in the loop containing the last pump to be started. For the first pump shutdown case, the transient is the reverse of the last pump startup transient. Design values for the pump starti ng/stopping conditions are given in Table 3.9(N)-1, along with the assumed number of occurrences. The 3800 occurrences listed in the table do not include the startups or shutdowns associated with RCS heatup and cooldown. 14. Turbine Roll Test This transient is imposed upon the plant during the hot functional test period for turbine cycle checkout. Reactor coolant pump power will be used to heat the reactor coolant to operating temperature (no-load conditions), and the steam generated will be used to perform a turbine roll test. However, the plant cooldown during this test will exceed the 100 F per hour design rate. The number of such test cycles is specified at 20 times, to be performed at the beginning of plant operating life prior to irradiation. 15. Primary Side Leakage Test Subsequent to each time the primary system has been opened, a leakage test will be performed. During this test, the primary system pressure is, for design purposes, raised to 2500 psia, with the system temperature above the minimum temperature imposed by reactor vessel material ductility requirements, while the system is checked for leaks. In actual practice, the primary system will be pressurized to approximately 2400 psig, as measured at the pressurizer , to prevent the pressurizer safety valves from lifting during the leak test.

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Page 12 During this leakage test, the secondary side of the steam generator must be pressurized so that the pressure diffe rential across the tube sheet does not exceed 1600 psi. This is accomplished with the steam, feedwater, and blowdown lines closed off. For design purposes, it is assumed that 200 cycles of this test will occur during the 40-year life of the plant. 16. Secondary Side Leakage Test During the life of the plant, it may be necessary to check the secondary side of the steam generator (particularly, the manway closure) for leakage.

For design purposes, it is assumed that the steam generator secondary side is pressurized to just below its de sign pressure, to prevent the safety valves from lifting. In order not to exceed a secondary side to primary side pressure differential of 670 psi, the primary side must also be pressurized. The primary system must be above the minimum temperature imposed by reactor vessel materials ductility requirements. It is assumed, that this test is performed 80 times during the 40-year life of the plant. 17. Tube Leakage Test During the life of the plant, it may be necessary to check the steam generator for tube leakage and tube to tube sheet leakage. This is done by visual inspection of the underside (channel head side) of the tube sheet for

water leakage, with the secondary side pressurized. Tube leakage tests are performed during plant cold shutdowns. For these tests, the secondary side of the steam generator is pressurized with water, initially at a relatively low pressure, and the primary system remains depressurized. The underside of the tube sheet is examined visually for leaks. If any are obs erved, the secondary side is then depressurized and repairs made by t ube plugging. The secondary side is then repressurized (to a higher pressu re), and the underside of the tube sheet is again checked for leaks. Th is process is repeated until all the leaks are repaired. The maximum (fin al) secondary side test pressure reached is 840 psig. The total number of tube leakage test cycles is defined as 800 during the 40-year life of the plant. Followi ng is a breakdown of the anticipated number of occurrences at each secondary side test pressure:

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Page 13 Test Pressure, psig Number Occurrences 200 400 400 200 600 120 840 80 Both the primary and secondary sides of the steam generators will be at the ambient temperatures during these tests. b. Upset Conditions The following primary system transients are considered upset conditions:

Loss of Load (without immediate reactor trip)

Loss of Power Partial Loss of Flow Reactor Trip from Full Power Inadvertent Reactor Coolant System Depressurization Inadvertent Startup of an Inactive Loop Control Rod Drop Inadvertent Safety Injection Actuation Operating Basis Earthquake Excessive Feedwater Flow RCS Cold Overpressurization S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 14 1. Loss of Load (Without Immediate Reactor Trip)

This transient applies to a step decrease in turbine load from full power (turbine trip) without immediately in itiating a reactor trip, and represents the most severe pressure transient on the RCS under upset conditions.

The reactor eventually trips as a consequence of a high pressurizer level

trip initiated by the Reactor Protection System (RPS). Since redundant means of tripping the reactor are provide d as a part of the RPS, transients of this nature are not expected, but are included to ensure a conservative design. The number of occurrences of this transient is specified at 80 times, or 2 times per year for the 40-year plant design life. 2. Loss of Power This transient applies to a blackout situation involving the loss of outside electrical power to the station, assumed to be operating initially at 100 percent power, followed by reactor and turbine trips. Under these circumstances, the reactor coolant pumps are de-energized and, following

coastdown of the reactor coolant pum ps, natural circulation builds up in the system to some equilibrium value. This condition permits removal of core residual heat through the steam generators which, at this time, are receiving feedwater, assumed to be at 32F, from the Emergency Feedwater System operating from diesel generator power. Steam is removed for reactor cooldown through atmospheric relief valves provided for this purpose. The number of occurrences of this transient is specified at 40 times or 1 per year for the 40-year plant design life.

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Page 15 3. Partial Loss of Flow This transient applies to a partial loss of flow from full power, in which a reactor coolant pump is tripped out of service as the re sult of a loss of power to that pump. The consequences of such an accident are a reactor and turbine trip, on low reactor coolant flow, followed by automatic opening of the Steam Dump System and fl ow reversal in the affected loop. The flow reversal causes reactor coolant at cold leg temperature to pass through the steam generator and be cooled still further. This cooler water

then flows through the hot leg piping and enters the reactor vessel outlet nozzles. The net result of the flow re versal is a sizable reduction in the hot leg coolant temperature of the affected loop. The number of occurrences of this transient is specified at 80 times, or 2 times per year for the 40-year plant design life. 4. Reactor Trip From Full Power A reactor trip from full power may occur from a variety of causes resulting in temperature and pressure transients in the RCS and in the secondary side of the steam generator.

This is the result of continued heat transfer from the reactor coolant in the steam generator. The transient continues until the reactor coolant and steam generator secondary side temperatures are in equilibrium at zero power conditions. A continued

supply of feedwater and controlled dumping of steam remove the core residual heat and prevent the steam generator safety valves from lifting. The reactor coolant temperature and pressure undergo a rapid decrease from full power values as the RPS causes the control rods to move into the core. Various moderator cooldown transients associated with reactor trips can occur as a result of excessive feed or steam dump after trip or large load increase. For design purposes, reactor trip is assumed to occur a total of 400 times or 10 times per year over the lif e of the plant. The various types of trips and the number of occurrences for each are as follows: (a) Reactor trip with no inadvertent cooldown - 230 occurrences (b) Reactor trip with cooldown but no safety injection - 160 occurrences S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 16 (c) Reactor trip with cooldown actuating safety injection -

10 occurrences. 5. Inadvertent Reactor Coolant System Depressurization Several events can be postulated as occurring during normal plant operation which will cause rapid depressurization of the RCS. These include: (a) Actuation of a single pressurizer safety valve (b) Inadvertent opening of one pressurizer power-operated relief valve due either to equipment malfunction or operator error (c) Malfunction of a single pressurizer pressure controller causing one power-operated relief valve and tw o pressurizer spray valves to open (d) Inadvertent opening of one pressu rizer spray valve, due either to equipment malfunction or operator error (e) Inadvertent auxiliary spray.

Of these events, the pressurizer safety valve actuation causes the most severe transient, and is used as an "umbrella" case to conservatively represent the reactor coolant pressure and temperature variations arising from any of them. When a pressurizer safety valve opens, and remains open, the system rapidly depressurizes, the reactor trips, and the Safety Injection System (SIS) is actuated. Also, the passive accumulators of the SIS are actuated when pressure decreases by approximately 1600 psi, about 12 minutes

after the depressurization begins.

The depressurization and cooldown are eventually terminated by operator action. All of these effects are completed within approximately 18 minutes. It is conservatively assumed

that none of the pressurizer heaters are energized. With pressure constant and safety in jection in operation, boiloff of hot leg liquid through the pressurizer and open safety valve will continue.

For design purposes, this transient is assumed to occur 20 times during the 40-year design life of the plant.

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Page 17 6. Inadvertent Startup of an Inactive Loop This transient can occur when a loop is out of service. With the plant operating at maximum allowable power level, the reactor coolant pump in the inactive loop is started as a resu lt of operator error. Reactor trip occurs on high nuclear flux. This transient is assumed to occur 10 times

during the life of the plant. 7. Control Rod Drop This transient occurs if a bank of control rods drops into the fully inserted position due to a single component failure. It is assumed that this transient occurs 80 times over the life of the plant. 8. Inadvertent Safety Injection Actuation A spurious safety injection signal results in an immediate reactor trip followed by actuation of the high head centrifugal charging pumps. These pumps deliver borated water from the RWST to the RCS cold legs. The

initial portion of this transient is similar to the Reactor Trip from Full Power with no cooldown. Controlled steam dump and feedwater flow after trip removes core residual heat. Reactor coolant temperature and pressure decrease as the control rods move into the core. Later in the transient, the inject ed water causes the RCS pressure to increase to the pressurizer power-ope rated relief valve setpoint and the primary and secondary temperatures to decrease gradually. The transient

continues until the operator stops the charging pumps. It is assumed that the plant is then returned to no-load conditions, with pressure and temperature changes controlled within normal limits. For design purposes, this transient is assumed to occur 60 times during the 40-year design life of the plant.

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Page 18 9. Operating Basis Earthquake The mechanical stresses resulting from the Operating Basis Earthquake are considered on a component basis. Fatigue analysis, where required by the codes, is performed by the supplier as part of the stress analysis report.

The earthquake loads are a part of the mechanical loading conditions specified in the equipment specifications. The origin of their determination is separate and distinct from those transients resulting from fluid pressure and temperature. Th ey are, however, considered in the design analysis. 10. Excessive Feedwater Flow An excessive feedwater flow transient is conservatively defined as an "umbrella" case to cover occurrence of several events of the same general nature. The postulated transient results from inadvertent opening of a

feedwater control valve while the plant is at the hot standby or no load condition, with the feedwater, condensate and heater drain systems in

operation. It is assumed, that the stem of a feedwater control valve fails and the valve immediately reaches the full open position. In the steam generator directly affected by the malfunc tioning valve ("failed loop"), the feedwater flow step increases from essentially zero flow to the value determined by the system resistan ce and the developed head of all operating feedwater pumps. Steam flow is assumed to remain at zero, and the temperature of the feedwater entering the steam generator is conservatively assumed to be 32 F. Feedwater flow is isolated on a reactor coolant low T avg signal; a low pressurizer pressure signal actuates the Safety Injection System. Emergency feedwater flow, initiated by the safety injection signal, is assumed to continue with all pumps discharging into the affected steam generator. It is assumed, for conservatism in the secondary side analysis, that there is emergency feedwater flow to the steam generators not affected by the malfunctioned valve in the "unfailed loops." Plant conditions stabilize at the values reached in 600 seconds, at which time emergency feedwater flow is terminated. The plant is then

either taken to cold shutdown, or re turned to the no load condition at a normal heatup rate with the Emergency Feedwater System under manual

control.

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Page 19 For design purposes, this transient is assumed to occur 30 times during the life of the plant.

11. RCS Cold Overpressurization RCS cold overpressurization occu rs during startup and shutdown conditions at low temperature, with or without the existence of a steam bubble in the pressurizer, and is especially severe when the reactor coolant system is in a water-solid configuration. The event is inadvertent, and can

potentially occur by any one of a variety of malfunctions or operator errors. All events, which have occurred to date, may be categorized as belonging to either events resulting in the addition of mass (mass input

transients) or events resulting in the addition of heat (heat input transients). All these possible transients are represented by composite, "umbrella" design transients, referred to here as RCS cold

overpressurization. The number of occurrences of this transient is specified at 10 times for the 40-year plant design operating period. c. Emergency Conditions The following primary system transients are considered Emergency Conditions:

Small Loss-of-Coolant Accident Small Steam Line Break Complete Loss of Flow.

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Page 20 1. Small Loss-of-Coolant Accident For design transient purposes, the sma ll loss-of-coolant accident is defined as a break equivalent to the severance of a 1-inch inside diameter branch connection. (Breaks smaller than 0.375 inch inside diameter can be handled by the normal makeup system, and produce no significant fluid systems transients.) Breaks which are much larger than 1 inch will cause

safety injection soon after the accident, and are regarded as Faulted Conditions. For design purposes, it is assumed that this transient occurs five times during the life of the plant. It is assumed that the Safety Injection System is actuated immedi ately after the break occurs, and subsequently delivers water at a minimum temperature of 32 F to the RCS. 2. Small Steam Line Break For design transient purposes, a small steam line break is defined as a break equivalent in effect to a steam safety valve opening and remaining open. This transient is assumed to occur five times during the life of the plant. The following conservative assumptions are used in defining the transients: (a) The reactor is initially in a hot, zero-power condition.

(b) The small steam line break results in immediate reactor trip and safety injection actuation. (c) A large shutdown margin, coupled with no feedback or decay heat, prevents heat generation during the transient. (d) The Safety Injection System operates at design capacity and repressurizes the RCS within a relatively short time. 3. Complete Loss of Flow This accident involves a complete loss of flow from full power, resulting from simultaneous loss of power to all reactor coolant pumps. The consequences of this incident are a reactor trip and turbine trip on undervoltage, followed by automatic opening of the Steam Dump System.

For design purposes, this transient is assumed to occur five times during the plant lifetime.

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Page 21 d. Faulted Conditions The following primary system transients are considered Faulted Conditions. Each of the following accidents should be evaluated for one occurrence: Reactor Coolant Pipe Break (Large Loss-of-Coolant Accident) Large Steam Line Break Feedwater Line Break Reactor Coolant Pump Locked Rotor Control Rod Ejection Steam Generator Tube Rupture Safe Shutdown Earthquake. 1. Reactor Coolant Pipe Break (Large Loss-of-Coolant Accident)

Following rupture of a reactor coolan t pipe resulting in a large loss of coolant, the primary system pressure decreases causing the primary system temperature to decrease.

Because of the rapid blowdown of coolant from the system and the comparatively large heat capacity of the metal sections of the components, it is likely that the metal will still be at or near the operating temperature by the end of blowdown. It is conservatively assumed that the Safety Injection System is actuated to introduce water at a minimum temperature of 32 F into the RCS. The safety injection signal will also result in reactor and turbine trips. 2. Large Steam Line Break This transient is based on the complete severance of the largest steam line. The following conservative assumptions were made: (a) The reactor is initially in a hot, zero-power condition.

(b) The large steam line break results in immediate reactor trip and in actuation of the Safety Injection Systems. (c) A large shutdown margin, coupled with no feedback or decay heat, prevents heat generation during the transients.

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Page 22 (d) The Safety Injection System operates at design capacity and repressurizes the Reactor Coolant System within a relatively short time. 3. Feedwater Line Break This accident involves a double-ended rupture of the main feedwater piping from full power, resulting in the rapid blowdown of one steam generator and the termination of main feedwater flow to the others. The blowdown is completed in approximately 27 seconds. Conditions were conservatively chosen to give the most severe primary side and secondary side transients. All emergency feedwater flow exits at the break. The incident is terminated when the operator manually realigns the Emergency Feedwater System to isolate the break and to deliver emergency feedwater to the intact steam generators. 4. Reactor Coolant Pump Locked Rotor This accident is based on the instantaneous seizure of a reactor coolant pump with the plant operating at full power. The locked rotor can occur in any loop. Reactor trip occurs almost immediately, as the result of low coolant flow in the affected loop. 5. Control Rod Ejection This accident is based on the single most reactive control rod being instantaneously ejected from the core. This reactivity insertion in a particular region of the core causes a severe pressure increase in the RCS such that the pressurizer safety valves will lift and also causes a more severe temperature transient in the l oop associated with the affected region than in the other loops. For conservatism the analysis is based on the

reactivity insertion and does not include the mitigating effects (on the

pressure transient) of coolant blowdown through the hole in the vessel

head vacated by the ejected rod.

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Page 23 6. Steam Generator Tube Rupture This accident postulates the double-ended rupture of a steam generator tube resulting in a decrease in pre ssurizer level and reactor coolant pressure. Reactor trip will occur due to the resulting safety injection signal. In addition, safety injection actuation automatically isolates the feedwater lines, by tripping all feedwater pumps and closing the feedwater isolation valves. When this accident occurs, some of the reactor coolant blows down into the affected steam ge nerator causing the shell side level to rise. The primary system pressure is reduced below the secondary

safety valve setting. Subsequent recovery procedures call for isolation of the steam line leading from the affected steam generator. This accident will result in a transient which is no more severe than that associated with a reactor trip from full power. Therefore, it requires no special treatment insofar as fatigue evaluation is concerned, and no specific number of

occurrences is postulated. 7. Safe Shutdown Earthquake The mechanical dynamic or static e quivalent loads due to the vibratory motion of the Safety Shutdown Earthquake are considered on a component basis. e. Test Conditions The following primary system transients under Test Conditions are discussed: Primary Side Hydrostatic Test Secondary Side Hydrostatic Test. 1. Primary Side Hydrostatic Test The pressure tests include both s hop and field hydrostatic tests which occur as a result of component or system testing. This hydro test is performed at a water temperature which is compatible with reactor vessel material ductility requirements and a test pressure of 3107 psig (1.25 times design pressure). In this test, the RCS is pressurized to 3107 psig, coincident with steam generator secondary side pressure of 0 psig. The RCS is designed for 10 cycles of these hydrostatic tests, which are performed prior to plant startup. The number of cycles is independent of other operating transients.

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Page 24 Additional hydrostatic tests will be performed to meet the in-service inspection requirements of ASME Section XI, Subarticle IS5-20. A total

of four such tests is expected. Th e increase in the fatigue usage factor caused by these tests is easily covered by the conservative number (200) of primary side leakage tests that are considered for design. 2. Secondary Side Hydrostatic Test The secondary side of the steam ge nerator is pressurized to 1.25 design pressure with a minimum water temperature of 120 F, coincident with the primary side at 0 psig. For design purposes, it is assumed that the steam generator will experience 10 cycles of this test.

These tests may be performed either pr ior to plant startup, or subsequently following shutdown for major repairs or both. 3.9(N).1.2 Computer Programs Used in Analyses The following computer programs have been used in dynamic and static analyses to determine mechanical loads, stresses, and deformations of seismic Category I components and equipment.

These are described and verified in References 1 and 2. a. WESTDYN Static and dynamic analysis of redundant piping systems

b. FIXFM-3 Time-history response of three-dimensional structures
c. WESDYN-2 Piping system stress analysis from time-history displacement data
d. STHRUST Hydraulic loads on loop components from blowdown information
e. WESAN Reactor coolant loop equipment support structures analysis and evaluation S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 25 f. WECAN Finite element structural analysis g. DARIWOSTAS Dynamic transient response analysis of reactor vessel and internals. 3.9(N).1.3 Experimental Stress Analysis No experimental stress analysis methods are used for seismic Category I systems or components. However, Westinghouse makes extensive use of measured results from prototype plants and various scale model tests, as discussed in Subsection 3.9(N).2.

3.9(N).1.4 Considerations for the Evaluation of the Faulted Condition

a. Loading Conditions The structural stress analyses performed on the Reactor Coolant System consider the loadings specified as shown in Table 3.9(N)-2. These loads result from thermal expansion, pressure, weight, Op erating Basis Earthquake (OBE), Safe Shutdown Earthquake (SSE), design basi s loss-of-coolant accident, and plant operational thermal and pressure transients. b. Analysis of the Reactor Coolant Loop The reactor coolant loop piping is evaluated in accordance with the Criteria of ASME III, NB-3650 and Appendix F. The loads included in the evaluation result from the SSE, dead weight, pressure, and LOCA loadings from the effect of the three branch nozzle breaks per Subsection 3.6(N).2.1 (loop hydraulic forces, asymmetric sub-compartment pressurization forces, and reactor vessel motion) and the secondary side breaks at the main steam line and main feedwater line terminal end nozzle locations at the steam generator. The loads used in the analysis of the Reactor Coolant Loop/Support System are described in detail below.

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Page 26 1. Pressure Pressure loading is identified as either membrane design pressure or general operating pressure, depe nding upon its application. The membrane design pressure is used in connection with the longitudinal pressure stress and minimum wall thickness calculations in accordance with the ASME Code. The term operating pressure is used in connection with determination of the system deflections and support forces. The steady-state operating hydraulic forces based on the system initial pressure are applied as general operating pressure loads to the reactor coolant loop model at changes in

direction or flow area. 2. Weight A weight analysis is performed to meet Code requirements by applying a 1.0 g load downward on the complete piping system. The piping is assigned a distributed mass or weight as a function of its properties. This method provides a distributed loading to the piping system as a function of the weight of the pipe and contained fluid during normal operating conditions. 3. Seismic The forcing functions for the reactor coolant loop piping seismic analyses are three orthogonal components of earthquake. The three components used to simulate the earthquake are in the form of statistically independent time-history accelerations. The ear thquake accelerations for the north-south direction and east-west direction are applied to the containment base mat in the respective global directions simultaneously

with the vertical acceleration along the vertical direction. For the OBE and SSE seismic analyses, damping values as defined in Section 3.7(N) are used in the reactor coolant loop/supports system

analysis.

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Page 27 4. Loss-of-Coolant Accident Blowdown loads are developed in the reactor coolant loop as a result of transient flow and pressure fluctua tions following a postulated pipe break in one of the three large RCL branch nozzles. Structural consideration of dynamic effects of postulated pipe br eaks requires postulation of a finite number of break locations. Postulated pipe break locations are given in Section 3.6(N). Time-history dynamic analysis is performed for these postulated break cases. Hydraulic models are used to generate time-dependent hydraulic

forcing functions used in the analysis of the reactor coolant loop for each break case. For a further descripti on of the hydraulic forcing functions, refer to Section 3.6(N).

5. Transients The Code requires satisfaction of certain requirements relative to operating transient conditions. Operating transients are discussed in

Subsection 3.9(N).1.1. The vertical thermal growth of the reactor pressure vessel nozzle centerlines is considered in the thermal analysis to account for equipment nozzle displacement as an external movement. The hot moduli of elasticity E, the coefficient of thermal expansion at the metal temperature , the external movements transmitted to the piping due to vessel growth, and the temperature rise above the ambient temperature T, define the required input data to perform the flexibility analysis for thermal expansion. To provide the necessary high degree of integrity for the Reactor Coolant System, the transient conditions selected for fatigue evaluation are based on conservative estimates of the magn itude and anticipated frequency of occurrence of the temperature and pressure transients resulting from various plant operating conditions.

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Page 28 c. Reactor Coolant Loop Analytical Models and Methods The analytical methods used in obtaini ng the solution consist of the transfer matrix method and stiffness matrix formula tion for the static structural analysis, time-history integration method for seismic dynamic analysis, and time-history integration method for the loss-of-coolant accident dynamic analysis. The integrated reactor coolant loop/supports system model is the basic system model used to compute loadings on components, component supports, and piping. The system model includes the stiffness and mass characteristics of the reactor coolant loop piping and components, the s tiffness of supports, and the stiffnesses of auxiliary line piping which affect the system. The deflection solution of the entire system is obtained for the various loading cases from which the internal member forces and piping stresses are calculated.

1. Static The reactor coolant loop/supports system model, constructed for the WESTDYN computer program, is repres ented by an ordered set of data which numerically describes the physical system. Figure 3.9(N)-1 shows an isometric line schematic of this mathematical model. The steam generator and reactor coolant pump vertical and lateral support members are described in Subsection 5.4.14. The spatial geometric description of the reactor coolant loop model is based upon the reactor coolant loop piping layout and equipment

drawings. The node point coordinates and incremental lengths of the members are determined from these drawings. Geometrical properties of

the piping and elbows along with the modulus of elasticity E, the coefficient of thermal expansion , the average temperature change from ambient temperature T, and the weight per unit length are specified for each element. The primary equipment supports are represented by stiffness matrices which define restraint characteristics of the supports.

Due to the symmetry of the static lo adings, the reactor pressure vessel centerline is represented by a fixed boundary in the system mathematical model. The vertical thermal growth of the reactor vessel nozzle centerline is considered in the construction of the model.

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Page 29 The model is made up of a number of sections, each having an overall transfer relationship formed from its group of elements. The linear elastic properties of the section are used to define the stiffness matrix for the section. Using the transfer relationshi p for a section, the loads required to suppress all deflections at the ends of the section arising from the thermal and boundary forces for the section are obtained. These loads are incorporated into the overall load vector. After all the sections have been defined in this matter, the overall stiffness matrix and associated load vector to suppress the deflection of all the network points is determined. By inverting the stiffness matrix, the flexibility matrix is determined. The flexibility matrix is multiplied by the negative of the load vector to determine the network point deflections due to the thermal and boundary force effects. Using the general transfer relationship, the deflections and internal forces are then determined at all node points in the system. The static solutions for weight, thermal, and general pressure loading conditions are obtained by using the WESTDYN computer program. The derivation of the hydraulic loads for the loss-of-coolant accident analysis of the loop is covered in Subsection 3.6(N).2. 2. Seismic The model used in the static analysis is modified for the dynamic analysis by including the mass characteristics of the piping and equipment. In the time-history seismic analysis, the containment internals structure and all

of the piping loops are included in the coupled system model. The effect of the equipment motion on the Reactor Coolant Loop/Supports System is obtained by modeling the mass and the stiffness characteristics of the equipment in the overall system model. The steam generator is represented by three discrete masses. The lowest mass is located at the intersection of the centerlines of the inlet and outlet nozzles of the steam generator. The middle mass is located at the steam generator upper support elevation and the third mass is located at the top of the steam generator. The reactor coolant pump is typically represented by a two discrete mass model. The lower mass is located at the intersection of the centerlines of the pump suction and discharge nozzles. The upper mass is located near the center of gravity of the motor.

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Page 30 The reactor vessel and core internals are represented by three discrete masses. The masses are lumped at various locations representing the pressure vessel and reactor internals. The component upper and lower latera l supports are inactive during plant heatup, cooldown and normal plant opera ting conditions. However, these restraints become active when the pl ant is at power and under the rapid motions of the reactor coolant loop components that occur from the dynamic loadings and are represented by stiffness matrices and/or individual tension or compression spring members in the dynamic model.

The analyses are performed at the full power condition. The total response is obtained directly by direct time integration of the equations of motion. The results of the time-history analysis are forces and displacements. The time-history displacement response is then used in computing support loads and in performing the reactor coolant loop piping stress evaluation. 3. Loss-of-Coolant Accident The mathematical model used in the static analyses is modified for the loss-of-coolant accident analyses by including the mass characteristics of the piping and primary equipment.

The natural frequencies and eigenvectors are determined from this. The time-history hydraulic forces at the node points are combined to obtain the forces and moments acti ng at the corresponding structural lumped-mass node points. The dynamic structural solution for the full power loss-of-coolant accident and steam line break is obtained by using a modified-predictor-corrector-integration technique and normal mode

theory. When elements of the system can be represented as single acting members (tension or compression members), they are considered as nonlinear elements, which are represented mathematically by the combination of a gap, a spring, and a viscous damper. The force in this nonlinear element is

treated as an externally applied force in the overall normal mode solution. Multiple nonlinear elements can be applied at the same node, if necessary.

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Page 31 The time-history solution is performed in program FIXFM3. The input to this subprogram consists of the natural frequencies, normal modes, applied forces and nonlinear elements. The natural frequencies and normal modes for the modified reactor coolant loop dynamic model are determined with the WESTDYN program. To properly simulate the release of the strain

energy in the pipe, the internal forces in the system at the postulated break location due to the initial steady-state hydraulic forces, thermal forces, and weight forces are determined. Th e release of the strain energy is accounted for by applying the negative of these internal forces as a step function loading. The initial conditi ons are equal to zero because the solution is only for the transient problem (the dynamic response of the system from the static equilibrium position). The time-history displacement solution of all dynamic degrees of freedom is obtained using program FIXFM3 and employing 4 percent critical damping. The loss-of-coolant accident displacements of the reactor vessel are applied in time-history form as input to the dynamic analysis of the reactor coolant loop. The loss-of-coolant accident analysis of the reactor vessel includes all the forces acting on the vessel including internals reactions and loop mechanical loads. The react or vessel analysis is described in Subsection 3.9(N).1.5. The resultant asymmetric external pressure loads on the RCP and steam generator resulting from a postulated pi pe rupture and pressure build-up in the loop compartments, are applied to the same integrated RCL/supports system model used to compute loadings on components, component

supports and RCL piping as discussed above. The response of the entire system is obtained for the various external pressure loading cases from which the internal member forces and piping stresses are calculated for each pipe break case considered. The equipment support loads and piping stresses resulting from the external pressure loading are added to the support loads and piping stresses calculated using the loop LOCA hydraulic forces and RPV motion. The break locations considered for subcompartment pressurization are those postulated for the RCL LOCA analysis, as discussed in Section 3.6(N) and WCAP-8172 (Reference 1 of Section 3.6(N)). The asymmetric subcompartment pressure loads are provided to Westinghouse

by United Engineers & Constructors Inc. The analysis to determine these loads is discussed in Section 6.2.

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Page 32 The time-history displacement response of the loop is used in computing support loads and in performing stress evaluation of the reactor coolant loop piping. The time-history displacements of the FIXFM3 (or WESTDYN) program are used as input to WESDYN-2 (or WESTDYN) to determine the internal forces, deflections, and stresses at each end of the piping elements. For this calculation the displacements are treated as imposed deflections on the reactor coolant loop masses.

4. Transients Operating transients in a nuclear power plant cause thermal and/or pressure fluctuations in the reactor coolant fluid. The thermal transients cause time-varying temperature distribu tions across the pipe wall. These temperature distributions resulting in pipe wall stresses may be further subdivided in accordance with the Code into three parts, a uniform, a linear, and a nonlinear portion. The uniform portion results in general

expansion loads. The linear portion causes a bending moment across the wall and the nonlinear portion causes a skin stress. The transients as defined in Subsection 3.9(N).1.1 are used to define the fluctuations in plant parameters. A one-dimensional finite difference heat conduction program is used to solve the thermal transient problem. The pipe is represented by at least fifty elements through the thickness of the pipe. The convective heat transfer coefficient employed in this program represents the time varying heat transfer due to free and forced convection. The outer surface is assume d to be adiabatic while the inner surface boundary experiences the temperature of the coolant fluid.

Fluctuations in the temperature of the coolant fluid produce a temperature distribution through the pipe wall thickness which varies with time. An arbitrary temperature distributi on across the wall is shown in Figure 3.9(N)-2. The average through-wall temperature, T A , is calculated by integrating the temperature distribution across the wall. This integration is performed for all time steps so that T A is determined as a function of time. dXtX,T H 1tT H O A S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 33 The range of temperature between the largest and smallest value of T A is used in the flexibility analysis to generate the moment loadings caused by the associated temperature changes. The thermal moment about the mi d-thickness of the wall caused by the temperature distribution through the wall is equal to: dXtX,T 2 HXEM H O The equivalent thermal moment produced by the linear thermal gradient as shown in Figure 3.9(N)-2 about the mid-wall thickness is equal to:

2 1 L H 12TEM Equating M L and M, the solution for T 1 as a function of time is: dXtX,T 2 H X H 12 tT H O 2 1 The maximum nonlinear thermal gradient, T 2, will occur on the inside surface and can be determined as the difference between the actual metal temperature on this surface and half of the average linear thermal gradient plus the average temperature. 2TtTtO,TtT 1 A 2I 5. Load Set Generation A load set is defined as a set of pressure loads, moment loads, and through-wall thermal effects at a given location and time in each transient. The method of load set generation is based on Reference 3. The through-wall thermal effects are functions of time and can be subdivided into four parts: (a) Average temperature (T A) is the average temperature through-wall of the pipe which contributes to general expansion loads.

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Page 34 (b) Radial linear thermal gradient which contributes to the through-wall bending moment (T 1). (c) Radial nonlinear thermal gradient (T 2) which contributes to a peak stress associated with shearing of the surface. (d) Discontinuity temperature (T A - T B) represents the difference in average temperature at the cross sections on each side of a discontinuity.

Each transient is described by at least two load sets representing the maximum and minimum stress stat e during each transient. The construction of the load sets is accomplished by combining the following to yield the maximum (minimum) stress state during each transient: (a) T 1 (b) T 2 (c) A T A - B T B (d) Moment loads due to T A (e) Pressure loads. This procedure produces at least twice as many load sets as transients for each point. For all possible load set combinations, the primary-plus-secondary and peak stress intensities, fatigue reduction factors and cumulative usage factors are calculated. The WESTDYN program is used to perform this

analysis in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB-3650. Since it is impossible to predict the order or occurrence of the transients over a forty-year life, it is assumed that the transients can occur in any se quence. This is a very conservative assumption.

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Page 35 The combination of load sets yi elding the highest alternating stress intensity range is used to calculate the incremental usage factor. The next most severe combination is then determined and the incremental usage factor calculated. This procedure is repeated until all combinations having allowable cycles <10 6 are formed. The total cumulative usage factor at a point is the summation of the incremental usage factors. d. Primary Component Supports Models and Methods The static and dynamic structural analyses employ the matrix method and normal mode theory for the solution of lumped-parameter, multimass structural models. The equipment support structure models are dual-purpose since they are required to: (1) quantitatively represent the elastic restraints which the supports impose upon the loop, and (2) evaluate the individual support member stresses due to the forces imposed upon the supports by the loop. A description of the supports is found in Subsection 5.4.14. Detailed models are developed using beam elements and plate elements, where applicable. The reactor vessel supports are modeled using the WECAN computer program. Structure geometry, topology and member properties are used in modeling. Steam generator and reactor coolant pump supports are modeled as linear or nonlinear springs. For each operating condition, the loads (obtained from the RCL analysis) acting on the support structures are appropriately combined. Reactor coolant loop normal and upset conditions thermal expansion loads are treated as primary loading for the primary component supports. The adequacy of each member of the steam generator supports, reactor coolant pump supports, and pressurizer supports is verified by solving the ASME III Subsection NF stress and interaction

equations. The adequacy of the RPV S upport Structure is verified using the WECAN computer program and comparing the resultant stresses to the criteria given in ASME III Subsection NF. The respective computer programs are used with these models to obtain support stiffness matrices and member influence coefficients for the steam generator, reactor coolant pump, pressurizer and reactor vessel supports. Unit force along and unit moment about each coordinate axis are applied to the models at the equipment vertical centerline joint. Stiffness analyses are performed for each unit load for each model.

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Page 36 Joint displacements for applied unit loads are formulated into flexibility matrices. These are inverted to obtain support stiffness matrices which were included in the reactor coolant loop model. Loads acting on the supports obtained from the reactor coolant loop analysis, support structure member properties, and influence coefficients at each end of each member are input into the WESAN program. For each support case used, the following is performed: 1. Combine the various types of s upport plane loads to obtain operating condition loads (Normal, Upset, Emergency or Faulted). 2. Multiply member influence coefficients by operating condition loads to obtain all member internal forces and moments. 3. Solve appropriate stress or interaction equations for the specified operating condition. Maximum normal stress, shear stress, and combined load interaction equation values are printed as a ratio of maximum actual values divided by limiting values.

ASME Boiler and Pressure Vessel Code Section III, Subsection NF, stress and interaction equations are used with limits for the operating condition specified. e. Analysis of Primary Components Equipment which serves as part of th e pressure boundary in the reactor coolant loop includes the steam generators, the reactor coolant pumps, the pressurizer, and the reactor vessel. This equipment is ANS Safety Class 1 and the pressure boundary meets the requirements of the AS ME Boiler and Pressure Vessel Code,Section III, Subsection NB. This equipment is evaluated for the loading combinations outlined in Table 3.9(N)-2. The equipment is analyzed for: (1) the normal loads of deadweight, pressure and thermal, (2) mechanical transients of OBE, SSE, and pipe ruptures, and (3) pressure and temperature transients outlined in Subsection 3.9(N).1.1.

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Page 37 The results of the reactor coolant loop analysis are used to determine the loads acting on the equipment nozzles and the support/component interface locations.

These loads are supplied for all loading conditions on an "umbrella" load basis.

That is, on the basis of previous plant analysis, a set of loads is determined which should be larger than those seen in any single plant analysis. The "umbrella" loads represent a conservative means of allowing detailed component analysis prior to the completion of the system analysis. Upon completion of the system analysis, conformance is demonstrated between the actual plant loads and the

loads used in the analyses of the com ponents. Any deviations where the actual load is larger than the "umbrella" load will be handled by individualized analysis. Seismic analyses are performed individually for the reactor coolant pump, the pressurizer, and the steam generator. Detailed and complex dynamic models are used for the dynamic analyses. The response spectra corresponding to the building elevation at the highest component/building attachment elevation is used for the component analysis. Seismic analyses for the steam generator and pressurizer are performed using 2 percent damping for the OBE and 4 percent damping for the SSE. The analysis of the reactor coolant pump for determination of loads on the motor, main flange, and pump internals is performed using the damping for bolted steel structures, that is, 4 percent for the OBE and 7 percent for the SSE (2 percent for OBE and 4 percent for SSE is used in the system analysis). This damping is applicable to the reactor coolant pump since the main flange, motor stand, and motor are all bolted assemblies (see Section 5.4). The reactor pressure vessel is qualified by sta tic stress analysis based on loads that have been derived from dynamic analysis. The pressure boundary portions of Class 1 valves in the RCS are designed and analyzed according to the requirements of NB-3500 of ASME III. Valves in sample lines connected to the RCS are not considered to be ANS Safety Class 1 nor ASME Class 1. This is b ecause the nozzles where the lines connect to the primary system piping are orificed to a 3/8 inch hole (see Section 5.4.3.2).

This hole restricts the flow so that loss through a severance of one of these lines can be made up by normal charging flow.

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Page 38 3.9(N).1.5 Dynamic Analysis of Reactor Pr essure Vessel for Postulated Loss-of-Coolant Accident a. Introduction This section presents the method of computing the reactor pressure vessel response to a postulated loss-of-coolant accide nt (LOCA). The structural analysis considers simultaneous application of the time-history loads on the reactor vessel resulting from the reactor coolant loop mechanical loads, and internal hydraulic pressure transients. The vessel is rest rained by reactor vessel support pads and

shoes beneath four of the reactor vessel no zzles and the reactor coolant loops with the primary supports of the steam generators and the reactor coolant pumps. b. Interface Information All input information was developed within Westinghouse. This information includes: reactor internals properties, loop mechanical loads and loop stiffnesses, internal hydraulic pressure transients, and reactor support stiffnesses. These inputs allowed formulation of the mathematical models performance of the analyses, as will be described. c. Loading Conditions Following a postulated pipe rupture at one of the three branch nozzle breaks per Subsection 3.6(N).2.3, the reactor vessel is excited by time-history forces. As previously mentioned, these forces are the combined effect of two phenomena: (1) reactor coolant loop mechanical loads, and (2) reactor internal hydraulic

forces. The reactor coolant loop mechanical forces are derived from the elastic analysis of the loop piping for the postulated break. This analysis is described in Subsection 3.9(N).1.4c. The loop mechani cal forces which are released at the broken nozzle are applied to the vessel in the RPV blowdown analysis.

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Page 39 The reactor internals hydraulic pressure transients were calculated with the assumption that the structural motion is coupl ed with the pressure transients. This phenomenon has been referred to as hydroe lastic coupling or fluid-structure interaction. The hydraulic analysis considers the fluid-structure interaction of the core barrel by accounting for the deflecti ons of constraining boundaries which are represented by masses and springs. The dynamic response of the core barrel in its beam bending mode responding to blowdown forces compensates for internal pressure variation by increasing the volume of the more highly pressurized regions. The analytical methods used to develop the reactor internals hydraulics are described in WCAP-8708 (Reference 9). d. Reactor Vessel and Internals Modeling The reactor vessel is restrained by two mechanisms: (1) the four attached reactor coolant loops with the steam generator and reactor coolant pump primary supports

and (2) four reactor vessel supports, two beneath reactor vessel inlet nozzles and two beneath reactor vessel outlet nozzl es. The reactor vessel supports are described in Subsection 5.4.14 and are shown in Figure 5.4-14, and Figure 3.8-26.

The support shoe provides restraint in th e horizontal directi ons and for downward reactor vessel motion. The reactor vessel model consists of two nonlinear elastic models connected at a common node. One model represents the dynamic vertical characteristics of the

vessel and its internals, and the other model represents the translational and rotational characteristics of the structure. These two models are combined in the DARIWOSTAS code (Reference 1) to represent motion of the reactor vessel and its internals in the plane of the vessel centerline and the broken pipe centerline. The model for horizontal motion consider s that each node has one translational and one rotational degree of freedom in the vertical plane containing the centerline of the nozzle attached to th e broken pipe and the centerline of the vessel. A combination of beam elements and concentrated masses are used to represent the components including the vesse l, core barrel, neutron panels, fuel assemblies, and upper support columns. Connections between the various components are either pin-pin rigid links, translational impact springs with damping, or rotational springs.

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Page 40 The model for vertical motion considers that each mass node has one translational degree of freedom. The structure is represented by concentrated masses, springs, dampers, gaps, and frictional elements. The model includes the core barrel, lower support columns, bottom nozzles, fuel r ods, top nozzles, upper support structure, and reactor vessel. The horizontal and vertical models are coupled at the elevation of the primary nozzle centerlines. Node 1 of the horizontal model is coupled with node 2 of the vertical model at the reactor vessel no zzle elevation. This coupled node has external restraints characterized by a 3x3 matrix which represents the reactor coolant loop stiffness characteristics, by linear horizontal springs which describe the tangential resistance of the supports , and by individual nonlinear vertical stiffness elements which provide downw ard restraint only. The individual supports are located at the actual support pad locations and accurately represent the independent nonlinear behavior of each support. e. Analytical Methods The time-history effects of the internals loads and loops mechanical loads are combined and applied simultaneously to the appropriate nodes of the mathematical model of the reactor vessel and internals. The analysis is performed by numerically integrating the differential equations of motion to obtain the

transient response. The output of the analysis includes the displacements of the reactor vessel and the loads in the reactor vessel supports which are combined

with other applicable faulted condition lo ads and subsequently used to calculate the stresses in the supports. Also, the reactor vessel displacements are applied as a time-history input to the dynamic reactor coolant loop blowdown analysis. The resulting loads and stresses in the piping components and supports include both loop blowdown loads and reactor vessel displacements. Thus, the effect of vessel displacements upon loop response and th e effect of loop blowdown upon vessel displacements are both evaluated.

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Page 41 f. Results of Analysis As described, the reactor vessel and inte rnals were analyzed for three postulated break locations. Table 3.9(N)-12 summarizes the allowable and no-loss-of-function displacements for reactor internals. Positive vertical displacement is up and positive horizontal displacement is away from and along the centerline of the vessel nozzle in the loop in which the break was postulated to occur. These displacements were calculated using an assumed break opening area for the postulated pipe ruptures of 144 in

2. Based on the deterministic fracture mechanics evaluation of the RCS loop piping, Westinghouse has demonstrated

that postulation of pipe ruptures in the RC loop need not be made. An exemption from a portion of the requirements of General Design Criterion 4 of Appendix A to 10 CFR Part 50 has been granted to Seabrook; see Reference 14. The result of postulating rupture of one of the three branch line nozzles would be to impose reduced asymmetric loadings on the Reactor Core System. The fuel assembly

grid load due to pipe ruptures was not applied to the analysis results. The loads from the two sources, the internals reactions and the loop mechanical forces, are applied simultaneously in a nonlinear elastic dynamic time-history analysis on the model of the vessel, supports and internals. The results of this analysis are the dynamic loads on th e reactor vessel supports and vessel time-history displacements. The maximum loads are combined with other applicable loads, such as seismic and deadweight and applied statically to the vessel support structure. The maximum st resses in the support are calculated and compared to faulted condition stress allowables given in Subsection 3.9(N).1.6. 3.9(N).1.6 Stress Criteria for Cla ss 1 Components and Component Supports All Class 1 components and supports are designed and analyzed for the Design, Normal, Upset, and Emergency Conditions to the rules and requirements of the ASME Code Section III. The design analysis or test methods and associated stress or load allowable limits that will be used in evaluation of Faulted Conditions are those that are defined in Appendix F of the ASME Code, as amended by R.G. 1.124 and variations stated in Updated FSAR Section 1.8, with supplementary options outlined below:

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Page 42 a. Elastic System Analysis and Component Inelastic Analysis This is an acceptable method of evaluati on for Faulted Conditions if the rules of F1323.1(a) are met for component supports, within the scope of Subsection NF and if primary stress limits for components are taken as greater of 0.70 S u or S y+1/3(S u-S y) for membrane stress and greater of 0.70 S ut or S y+ 1/3(S ut-S y) for membrane-plus-bending stress, where material properties are taken at appropriate temperature. If plastic component analysis is used with elastic system analysis or with plastic system analysis, the deformations and displacements of the individual system members will be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis. b. Elastic/Inelastic System Analysis and Component/Test Load Method The test load method given in F-1370(d) is an acceptable method of qualifying components in lieu of satisfying the stress/load limits established for the component analysis. If the component/test load method is used with elastic or plastic system analysis, the deformations and displacements of the individual component members taken from the test load method data at the loads resulting from the system analysis will be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis. Loading combinations and allowable stresses for ASME III Class 1 components and supports are given in Table 3.9(N)-2 and Table 3.9(N)-3. Fo r Faulted condition evaluations, the effects of the safe shutdown earthquake (SSE) and loss-of-coolant accident (LOCA) are combined using the square-root-of-the-sum-of-the-squares (SRSS) method. Justification for this method of load combinations is contained in References 4 and 5. 3.9(N).1.7 Analytical Methods for RCS Class 1 Branch Lines For faulted conditions analysis of Class 1 branch piping attached to the reactor coolant loop, Equation (9) of ASME III Subsection NB-3652 is applied with a stress limit of 3.0 S

m. This criterion provides sufficient assurance that the piping will not collapse or experience gross distortion such that the function of the system would be impaired. The basis for this position is described in Westinghouse response to NRC Question 110.34 on the RESAR-414 application (Docket No. STN 50-572), which subsequently received a preliminary design approval (PDA) in November 1978.

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Page 43 The analytical methods used to obtain the solution consist of the transfer matrix method and stiffness matrix formulation for the static structural analysis, the response spectrum method for seismic dynamic analysis, and dynamic structural anal ysis for the effect of the three RCL branch nozzle breaks per Subsection 3.6(N).2.3. The integrated Class 1 piping/supports system model is the basic system model used to compute loadings on components, component and piping supports, and piping. The system models include the stiffness and mass characteristics of the Class 1 piping components, the reactor coolant loop, and the stiffness of supports which affect the system response. The deflection solution of the entire system is obtained for the various loading cases from which the internal member forces and piping stresses are calculated.

a. Static The Class 1 piping system models are constructed for the WESTDYN computer program, which numerically describes the physical system. A network model is made up of a number of sections, each having an overall transfer relationship formed from its group of elements. The lin ear elastic properties of the section are used to define the characteristics stiffness matrix for the section. Using the transfer relationship for a section, the lo ads required to suppress all deflections at the ends of the section arising from the thermal and boundary forces for the

section are obtained. After all the sections have been defined in this manner, the overall stiffness matrix and associated load vector to suppress the deflection of all the network points is determined. By inverting the stiffness matrix, the flexibility matrix is determined. The flexibility matrix is multiplied by the negative of the load vector to determine the network point deflections due to the thermal and boundary force effects. Using the general transfer relati onship, the deflections and internal forces are then determined at all node points in the system. The support loads are also computed by multiplying the stiffness matrix by the displacement vector at the support point. b. Seismic The models used in the static analyses are modified for use in the dynamic analyses by including the mass characteristics of the piping and equipment.

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Page 44 The lumping of the distributed mass of the piping systems is accomplished by locating the total mass at points in the system which will appropriately represent the response of the distributed system. Effects of the primary equipment motion, that is, reactor vessel, steam generator, reactor coolant pump, and pressurizer, on the Class 1 piping system are obtained by modeling the mass and the stiffness characteristics of the primary equipment and loop piping in the overall system model. The supports are represented by stiffness matrices in the system model for the dynamic analysis. Shock suppressors which resist rapid motions are also included in the analysis. The solution for the seismic disturbance employs the response spectra method. This method employs the lumped mass technique, linear elastic properties, and the principle of modal superposition. The total response obtained from the seismi c analysis consists of two parts: the inertia response of the piping system and the response from differential anchor motions. The stresses resulting from the anchor motions are considered to be secondary and, therefore, are included in the fatigue evaluation. c. Loss-of-Coolant Accident The mathematic models used in the seismic analyses of the Class 1 lines are also used for three RCL branch nozzle break effect analyses. To obtain the response

for lines attached to the unaffected loops and lines attached to the affected loop, the deflections from the analysis of the reactor coolant loop are applied at branch nozzle connections.

d. Fatigue A thermal transient heat transfer analysis is performed for each different piping component on all the Class 1 branch lines. The normal, upset, and test condition transients identified in Subsection 3.9.1.1 are considered in the fatigue evaluation. The thermal quantities T 1 , T 2 , and a a T a , -a b T b are calculated on a time-history basis, using a one-dimensional finite difference heat transfer computer program.

Stresses due to these quantities were calculated for each time increment using the methods of NB-3650 of ASME III. For each thermal transient, two loadsets are defined, representing the maximum and minimum stress states for that transient.

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Page 45 As a result of the normal mode spectral technique employed in the seismic analysis, the load components cannot be gi ven signed values. Eight load sets are used to represent all possible sign permutations of the seismic moments at each point, thus insuring the most conservative combinations of seismic loads are used in the stress evaluation. The WESTDYN computer program is used to calculate the primary-plus-secondary and peak stress in tensity ranges, fatigue reduction factors and cumulative usage factors for all possible load set combinations. It is conservatively assumed that the transien ts can occur in any sequence, thus resulting in the most conservative and restrictive combinations of transients. The combination of load sets yielding th e highest alternating stress intensity range is determined and the incremental usage factor calculated. Likewise, the next most severe combination is then determined and the incremental usage factor calculated. This procedure is repeated until all combinations having allowable cycles <10 6 are formed. The total cumulative usage factor at a point is the summation of the incremental usage factors. 3.9(N).2 Dynamic Testing and Analysis 3.9(N).2.1 Preoperational Vibration and Dynamic Effects Testing on Piping A preoperational piping vibrational and dynamics effects testing program will be conducted for the Reactor Coolant Loop/Supports System during pre-operational testing. The purpose of these tests will be to confirm that the system has been adequately designed and supported for vibration, as required by Section III of the AS ME Code, paragraph NB-3622.3. The tests will include reactor coolant pump starts and trips. If vibration is experienced, which, from visual observation, appears to be excessive, either: (1) an instrumented test program on the piping, will be conducted and the system reanalyzed to dem onstrate that the observed levels will not cause ASME Code stress and fatigue limits to be exceed ed, (2) the cause of the excessive vibration will be eliminated, or (3) the support system will be modified to reduce the vibration. Particular attention will be provided at those locations where the vibration is expected to be the most

severe for the particular transient condition being studied.

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Page 46 It should be noted that the layout, size, etc., of the reactor coolant loop and surge line piping used in the Seabrook plant is very similar to that employed in Westinghouse plants now in operation. The operating experience that has been obtained from these plants indicates that the reactor coolant loop and surge line piping are adequately designed and supported to minimize vibration. In addition, vibration levels of the reactor coolant pump, which is the only mechanical component that could cause vibration of the reactor coolant loop and surge line piping, are monitored as described in Subsection 5.4.1. Thus , excessive vibration of the reactor coolant loop and surge line piping should not be present. Howe ver, as added assurance that excessive piping vibration will not be present, the piping adjacent to the reactor coolant pump will be subjected to visual observation, as discussed above. 3.9(N).2.2 Seismic Qualification Testing of Safety-Related Mechanical Equipm ent The operability of seismic Category I mechanical equipment must be demonstrated if the equipment is active, i.e., ASME Section III components that must perform a mechanical motion during the course of performing their safety function in shutting down the plant to and maintaining it in a cold shutdown or in mitigating the consequences of a postulated event. The operability of active Safety Class 2 and 3 pumps, active Safety Class 1, 2 or 3 valves and their respective drives, operators and vital auxiliary equipment is shown by satisfying the criteria given in Subsection 3.9(N).3.2. Other active mechanical equipment is shown operable by either testing, analysis or a combination of testing and analysis. The operability programs implemented on this other active equipment are similar to the program described in Subsection 3.9(N).3.2 for pumps and valves. Testing procedures similar to the procedures outlined in Section 3.10(N) for electrical equipment are 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 amenable to modeling and dynamic analysis. Inactive seismic Category I equipment such as heat exchangers, racks and consoles are shown to have structural integrity during a seismic event by one of the following methods: (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 test environment.

As indicated above, Westinghouse used analysis, test, or a combination of test and analysis to seismically qualify originally-designated active equipment. Testing was the preferred method.

However, analysis was used when one of the following conditions was satisfied: a. The equipment is too large or the external loads, connecting elements, or appurtenances cannot be simulated with a shaker table test. b. The only requirement that must be satisfied relative to the safety of the plant is the maintenance of structural integrity (mechanical equipment only).

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Page 47 c. The component represents a simple linear system, or nonlinearities can be conservatively accounted for in the analysis. The seismic qualification of electrical equipment by analysis was not employed by Westinghouse. However, analysis was employed to supplement tests or to provide verification that test results are applicable for a particular configuration. Components not originally-designated as active were reviewed for loading and seismic accelerations to ensure that they meet active component requirements. A list of seismic Category I equipment is provided in Table 3.2-1 and Table 3.2-2. 3.9(N).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 pressu res 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. The determination of the

forcing functions as a direct correlation of pressure oscillations cannot be practically performed independently of the dynamic characteristics of the structure. The main objective 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 fo rcing function can be deduced. These studies indicated that the most important forcing functions are flow turbulence, and pump-related excitation. The relevance of such excitation depends on many factors, such as type and location of component and flow conditions. The effects of these forcing functions have been studied from tests performed on models and prototype plants, as well as component tests (References 6, 7 and 8). The Indian Point No. 2 plant has been establishe d as the prototype for a four-loop plant internals verification program, and was fully instrumented and tested during hot functional testing. In addition, the Trojan plant instrumentation program provides, and the Sequoyah No. 1 instrumentation program will provide, prototype data applicable to Seabrook (References 7 and 8). Seabrook Station is similar to Indian Point No. 2; the only significant differences are the modifications resulting from the use of 17x17 fuel, replacement of the annular thermal shield with neutron shielding panels, and the change to the UHI-style inverted top hat support structure configuration. These differences are addressed below.

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Page 48 a. 17x17 Fuel The only structural changes in the internals resulting from the design change from the 15x15 to the 17x17 fuel assembly are th e guide tube and control rod drive line. The new 17x17 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 15x15 fuel assembly vibration characteristics. b. Neutron Shielding Pads Lower Internals The primary cause of core barrel excita tion is flow turbulence generated in the downcomer annulus (Reference 7). The vibration levels due to core barrel excitation for Trojan and Seabrook, bot h having neutron shielding pads, are expected to be similar. Since Seab rook 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 smaller power) (Reference 6). However, scale model test results (Reference 6) and results from Trojan (Reference 8) show th at 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 No. 2 (thermal shield configuration), lead to the conclusion that stresses approximately equal to those of Indian Point No. 2 will resu lt on the Seabrook internals, with the attendant large safety margins. c. UHI-Style Inverted Top Ha t Upper Support Configuration The components of the upper internals are excited by turbulent forces due to axial and cross flows in the upper plenum (Reference 7). Sequoyah and Seabrook have the same basic upper internals configura tion, therefore, the general vibration behavior is not changed. Since Seab rook has slightly higher velocities than Sequoyah, the Seabrook upper internals vibr ation is expected to be slightly greater than that for Sequoyah. However, scale-model test results and analytical work predict high safety margins for the Sequoyah upper internals components.

Therefore, the Sequoyah results can be used quite confidently to demonstrate the adequacy of the Seabrook upper internals. The original test and analysis of the four-loop configuration is augmented (References 6, 7 and 8) to cover the effects of successive hardware modifications.

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Page 49 3.9(N).2.4 Preoperational Flow-Induced Vi bration Testing of Reactor Internals Because the Seabrook reactor internals design configuration is well characterized, as was discussed in Subsection 3.9(N).2.3, it is not considered necessary to conduct instrumented tests of the Seabrook plant hardware. The recommendations of Regulatory Guide 1.20 will be met by conducting the confirmatory preoperational testing examination for integrity. This examination will include approximately 30 features, with special emphasis on the following areas: a. All major load-bearing elements of the reactor internals relied upon to retain the core structure in place b. The lateral, vertical and torsional restraints provided within the vessel c. Those locking and bolting devices whos e failure could adversely affect the structural integrity of the internals d. Those other locations on the reactor internal components which are similar to those which were examined on the prototype designs e. 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 of foreign material are in evidence. A particularly close inspection will be made on the following items or areas using a 5X or 10X magnifying glass, or other appropriate inspection. a. Lower Internals

1. Upper barrel to flange girth weld.
2. Upper barrel to lower barrel girth weld.
3. Upper core plate aligning pin. Examine bearing surfaces for shadow marks, burnishing, buffing or scoring. Inspect welds for integrity. 4. Irradiation specimen guide screw locking devices and dowel pins. Check for lockweld integrity. 5. Baffle assembly locking devices. Check for lockweld integrity.
6. Lower barrel to core support girth weld.

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Page 50 7. Neutron shielding pad screw locking devices and dowel pin lock welds. Examine the interface surfaces for evidence of tightness. Check for lockweld integrity. 8. Radial support key welds. 9. Insert screw locking devices. Examine soundness of lockwelds.

10. Core support columns and instrumenta tion guide tubes. Check the joints for tightness and soundness of the locking devices. 11. Secondary core support assembly screw locking devices for lockweld integrity. 12. Lower radial support keys and inserts. Examine bearing surfaces for shadow marks, burnishing, buffing or scoring. Check the integrity of the lockwelds. 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 tw
o. Subsequent to the hot functional testing, the bearing surfaces of the key and keyway will show burnishing, buffing or shadow marks which indi cate pressure loading and relative motion between these parts. Minor sc oring of engaging surfaces is also possible and acceptable. 13. Gaps at baffle joints. Check gaps between baffle to baffle joints. b. Upper Internals
1. Guide tube, support column and orifice plate locking devices. 2. Support column welds.
3. Upper core plate alignment inserts. Examine for shadow marks, burnishing, buffing or scoring. Check the locking devices for integrity of lockwelds. 4. Guide tube enclosure welds, tube-transition plate welds and card welds. Acceptance standards are the same as required in the shop by the original design drawings and specifications.

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Page 51 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 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />. This provides a cyclic loading of approximately 10 7 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.

Pre- and post-hot-functional inspection results serve to confirm that the internals are well behaved. When no signs of abnormal wear or harmful vibrations are detected, and no apparent structural changes take place, the four-loop co re support structures are considered to be structurally adequate and sound for operation. 3.9(N).2.5 Dynamic System Analysis of the Reactor Internals under Faulted Conditions Analysis of the reactor internals for blowdown loads resulting from a loss-of-coolant accident is based on the time-history response of the internals to simultaneously applied blowdown forcing functions. The forcing functions are defined at points in the system where changes in cross section or direction of flow occur, such that differential loads are generated during the blowdown transient. The dynamic mechanical analysis can employ the displacement method, lumped parameters, stiffness matrix formulations, and assumes that all components behave in a linearly elastic manner. In addition, because of the complexity of the system and the components, it is necessary to use finite element stress analysis codes to provide more detailed information at various points.

A computer program, MULTIFLEX (Reference 9 and 10), which was developed to calculate local fluid pressure, flow, and density transients that occur in pressurized water reactor coolant systems during a loss-of-coolant accident is applie d to the subcooled, transition, and saturated two-phase blowdown regimes. This MULTIFLEX code is based on the method of characteristics wherein the resulting set of ordinary differential equations, obtained from the laws of conservation of mass, momentum, and energy are solved numerically using a fixed mesh in both space and time.

Although spatially one-dimensional conservation laws are employed, the code can be applied to describe three-dimensional system geometries by use of the equivalent piping networks. Such piping networks may contain any number of pipes or channels of various diameters, dead ends, branches (with up to six pipes connected to each branch), contractions, expansions, orifices, pumps and free surfaces (such as in the pressurizer). System losses such as friction, contraction, expansion, etc., are considered.

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Page 52 The reanalysis performed in support of increased core power has made use of the MULTIFLEX (Reference 9 and 10) computer code. MULTIFLEX is an extension of the BLODWN-2 computer code and includes mechanical structure models and their interactions with the thermal-hydraulic system. Both versions of the MULTIFLEX code share a common hydraulic modeling scheme, with differences being confined to a more realistic downcomer hydraulic network and a more realistic core barrel structural model that accounts for non-linear boundary conditions and vessel motion. Generally, this improved modeling results in lower, more realistic, but still conservative hydraulic forces on the core barrel. The N RC staff has accepted (Reference 15) the use of MULTIFLEX 3.0 for calculating the hydraulic for ces on reactor vessel internals (Reference 16).

The MULTIFLEX code evaluates the pressure and velocity transients throughout the system.

These pressure and velocity transients are stored as a file and are made available to the programs LATFORC and FORCE2, which use detailed geometri c descriptions to calculate the loadings on the reactor internals.

The LATFORC code (Appendix A of Reference 9) computes the horizontal or lateral forces on the reactor vessel shell interior, the core barrel, and the neutron pads due to the pressure on the location at up to ten elevations.

The FORCE2 code (Appendix B of Reference 9) computes the vertical forces on the reactor internals components. Each reactor component for which FORCE2 calculations are required is designated as an element and assigned an element number. Forces acting on each of the elements are calculated by summing the effects of:

a. The pressure differential across the element.
b. Flow stagnation on, and unrecovered orifice losses across the element.
c. Friction losses along the element.

Input to the code, in addition to the blowdown pressure and velocity transients, includes the effective area of each element on which the force acts due to the pressure differential across the element, a coefficient to account for flow stagna tion and unrecovered orifice losses, and the total area of the element along which shear forces act.

The mechanical analysis (Reference 11) has been performed using conservative assumptions. Some of the more significant assumptions are: a. The mechanical and hydraulic analys es have considered the effect of hydroelasticity.

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Page 53 b. The reactor internals are represented by a multi-mass system connected with springs and dashpots simulating the elastic response and the viscous damping of the components. The modeling is conducted in such a way that uniform masses are lumped into easily identifiable discrete masses while elastic elements are represented by springs. The model described is considered to have a sufficient number of degrees of freedom to represent the most important modes of vibration in the vertical direction. This model is conservative in the sense that further mass-spring resolution of the system would lead to further attenuation of the shock effects obtained with the present model.

The pressure waves generated within the reactor are highly dependent on the location and nature of the postulated pipe failure. In general, the more rapid the severance of the pipe, the more severe the imposed loadings on the components. A one millisecond severance time is taken as the limiting case.

In the case of the hot leg branch line break, the vertical hydraulic forces produce an initial upward lift of the core. A raref action wave propagates through the reactor hot leg nozzle into the interior of the upper core barrel. Since the wave has not reached the flow annulus on the outside of the barrel, the upper barrel is subjected to an impulsive compressive wave. Thus, dynamic instability (buckling) or large deflections of the upper core barrel, or both, is a possible response of the barrel during a hot leg branch line break, and results in transverse loading on the upper core components as the fluid exits the hot leg nozzle.

In the case of the cold leg bran ch line break, a rarefaction wave propagates along a reactor inlet pipe, arriving first at the core barrel at the in let nozzle of the broken loop. The upper barrel is then subjected to a non-axisymmetric expansion radial impulse which changes as the rarefaction wave propagates both around the barrel and down the outer flow annulus between vessel and barrel. After the cold leg branch line break, th e initial steady-state hydraulic lift forces (upward) decrease rapidly (within a few milliseconds) and then increase in the downward direction. These cause the reactor core and lower support structure to move initially downward. If a simultaneous seismic event with the intensity of the SSE is postulated with the loss-of-coolant accident, the imposed loading on the internals component is additive and, therefore, the combined loading is considered even though the loading imposed by the earthquake is generally small compared to the blowdown loading.

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Page 54 The summary of the mechanical analysis follows: a. Mathematical Model of the Reactor Pressure Vessel and Analytical Method The mathematical model of the RPV is a three-dimensional nonlinear finite element model which represents the dynami c characteristics of the reactor vessel and its internals in the six geometric degrees of freedom. The model was developed using the WECAN computer code. The model consists of three concentric structural submodels connected by nonlinear impact elements and stiffness matrices. The first submodel represents the reactor vessel shell and associated components. The reactor vesse l is restrained by four reactor vessel

supports (situated beneath alternate nozzles) and by the attached primary coolant piping. Each reactor vessel support is m odeled by a linear horizontal stiffness and a vertical spring-gap element. The attached piping is represented by a stiffness matrix. The second submodel represents the reactor core barrel (RCB), neutron panels, lower support plate, tie plates, and secondary core support components. This submodel is physically located inside the first, and is connected to it by a stiffness matrix at the internals support ledge. Core barrel to vessel shell impact is represented by nonlinear elements at the core barrel flange, core barrel nozzle, and lower radial support locations. The third and innermost submodel represents the upper support plate, guide tubes, upper support columns, upper and lower co re plates, and fuel. The third submodel is connected to the first and second by stiffness matrices and nonlinear elements. Fluid-structure or hydro-elastic interac tion is included in the reactor pressure vessel model for seismic evaluation. The horizontal hydro-elastic interaction is

significant in the cylindrical fluid flow region between the core barrel and reactor vessel (the downcomer). Mass matrices with off-diagonal terms (horizontal degrees-of-freedom only) attach between nodes on the core barrel and reactor vessel shell. The diagonal terms of the mass matrix are similar to the lumping of water mass to the vessel shell 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.

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Page 55 The matrices are a function of the properties of two cylinders with a fluid in the cylindrical annulus, specifically; inside a nd outside radius of the annulus, density of the fluid and length of the cylinders. Vertical segmentation of the RCB allows inclusion of radii variations along the RCB height and approximates the effects of RCB beam deformation. These mass matrices were inserted between selected nodes on the core barrel and reactor vessel shell. For LOCA the time-history effects of the internals loads and loop mechanical loads are combined and applied simultane ously to the appropriate nodes of the mathematical model of the reactor vessel and internals. The analysis is performed by numerically integrating the differential equations of motion to obtain the

transient response. The output of the analysis includes the displacements of the reactor vessel and the reactor internals. b. Transverse Excitation Model for Blowdown Various reactor internal components are subjected to transverse excitation during blowdown. Specifically, the barrel, guide tubes, and upper support columns are analyzed to determine their response to this excitation. c. Core Barrel For the hydraulic analysis of the pressure transients during hot leg blowdown, the maximum pressure drop across the barrel is a uniform radial compressive impulse. The barrel is then analyzed for dynamic buckling using the following conservative assumptions: 1. The effect of the fluid environment is neglected

2. The shell is treated as simply supported. During cold leg blowdown, the upper barrel is subjected to a non-axisymmetric expansion radial impulse which changes as the rarefaction wave propagates both around the barrel and down the outer flow annulus between vessel and barrel.

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Page 56 The analysis of transverse barrel response to cold leg blowdown is performed as follows: 1. The core barrel is treated as a simply supported cylindrical shell of constant thickness between the upper flange weldment and the lower support plate weldment. No credit is taken for the supports at the barrel midspan offered by the outlet nozzles. This assumption leads to conservative deflection estimates of the upper core barrel. 2. The core barrel is analyzed as a shell with two variable sections to model the support flange and core barrel. 3. The barrel with the core and neutron shielding pads is analyzed as a beam fixed at the top and elastically suppor ted at the lower radial support and the dynamic response is obtained. d. Guide Tubes The guide tubes in closest proximity to the outlet nozzle of th e affected loop are the most severely loaded during a blowdow

n. The transverse guide tube forces decrease with increasing distance from the ruptured nozzle location. All of the guide tubes are designed to main tain the function of the control rods for a break size of 144 in 2 and smaller. No credit for the function of the control rods is assumed for break size areas above 144 in
2. However, the design of the guide tube will permit control rod operation in all but four control rod positions, which is sufficient to maintain the core in a s ubcritical configuration, for break sizes up to a double-ended hot leg break. This double-ended hot leg break imposes the limiting lateral guide tube loading. e. Upper Support Columns Upper support columns located close to the broken nozzle during hot leg branch line break will be subjected to transverse loads due to cross flow. The loads applied to the columns are computed with a method similar to the one used for the

guide tubes, i.e., by taking into consid eration the increase in flow across the column during the accident. The columns are studied as beams with variable section, and the resulting stresses are obtained using the reduced section modulus and appropriate stress risers for the various sections.

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Page 57 The effects of the gaps that could exis t between vessel and barrel, between fuel assemblies, and between fuel assemblies a nd baffle plates are considered in the analysis. Linear analysis will not provide information about the impact forces generated when components impinge each other, but can, and is, applied prior to gap closure. Reference 11 provides further details of the blowdown method used

in the analysis of the reactor internals. The stresses due to the safe shutdow n earthquake (vertical and horizontal components) are combined with the bl owdown stresses to obtain the largest principal stress and deflection. f. Results All reactor internals components were found to be within acceptable stress and deflection limits for both hot leg and cold leg loss-of-coolant accidents, occurring simultaneously with the safe shutdown earthquake. The results obtained from the linear anal ysis indicate that during blowdown, the relative displacement between the components will close the gaps and, consequently, the structures will impinge on each other, making the linear

analysis unrealistic and forcing the application of nonlinear methods to study the problem. Although linear analysis will not provide information about the impact forces generated when components impinge on each other, it can, and is, applied

prior to gap closure. The effects of th e gaps that could exist between vessel and barrel, between fuel assemblies, between fuel assemblies and baffle plates, and

between the control rods and their guide paths were considered in the analysis. Both static and dynamic stress intensities are within acceptable limits. In addition, the cumulative fatigue usage factor is also within the allowable usage factor of unity. The stresses due to the Safe Shutdow n Earthquake (vertical and horizontal components) were combined in the most unfavorable manner with the blowdown stresses to obtain the largest principal stress and deflection. These results indicate that the maximum deflections and stresses in the critical structures are below the established allowable limits. For the transverse

excitation, it is shown that the upper ba rrel does not buckle during a hot leg break, and that it has an allowable stress distribution during a cold leg break.

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Page 58 Even though control rod insertion is not re quired for plant shutdown, this analysis shows that most of the guide tubes will deform within the limits established experimentally to assure control rod in sertion. For the guide tubes deflected above the no-loss-of-function limit, it must be assumed that the rods will not drop.

However, the core will still shut down due to the negative reactivity insertion in the form of core voiding. Shutdown will be aided by the great majority of rods that do drop. Seismic deflections of the guide tubes are generally negligible by comparison with the no-loss-of-function limit. 3.9(N).2.6 Correlations of Reactor Internals Vibration Tests with the Analytical Results As stated in Subsection 3.9(N).2.3, it is not considered necessary to conduct instrumented tests of the Seabrook reactor vessel internals. Adequacy of these internals will be verified by use of the Sequoyah and Trojan results. References 7 and 8 describe predicted vibration behavior based on studies performed prior to the plant tests. These studies, which use analytical models, scale model test results, component tests, and results of previous plant tests, are used to characterize the forcing functions and establish component struct ural characteristics, so that the flow induced vibratory behavior and response levels for Seabrook are estimated. These estimates are then compared to values deduced from plant test data obtained from the Sequoyah and the Trojan internals vibration measurement programs. 3.9(N).3 ASME Code Class 1, 2 and 3 Components, Component Supports and Core Support Structures The ASME Code Class components are construc ted in accordance with the ASME Boiler and Pressure Vessel Code,Section III. Detailed discussion of ASME Code Class 1 components is provided in Section 5.4 and Subsection 3.9(N).1. For core support struct ures, design loading conditions are given in Subsection 3.9(N).5. In general, for reactor internals components a nd for core support structures, the criteria for acceptability in regard to mechanical integrity anal yses 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 rema ins 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.

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Page 59 For the loss of coolant plus the safe shutdow n earthquake 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: a. Following the loss-of-coolant accident, the functional criterion to be met for the reactor internals is that the plant can be shutdown and cooled in an orderly fashion, so that fuel cladding temperature is kept within specified limits. This criterion implies that the deformation of critical components must be kept sufficiently small to allow core cooling. b. For large reactor coolant branch nozzle pipe breaks, the reduction in water density greatly reduces the reactivity of the core, thereby shutting down the core whether

the control 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 is to assure effectiveness of the Emergency Core Cooling System. Inse rtion 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. c. The inward upper barrel deflections are controlled to insure no contacting of the nearest rod cluster control guide tube.

The outward upper barrel deflections are controlled in order to maintain an ade quate annulus for the coolant between the vessel inner diameter and core barrel outer diameter. d. The rod cluster control guide tube deflections are limited to insure operability of the control rods. e. To insure no column loading of rod cluster control guide tubes, the upper core plate deflection is limited.

Methods of analysis and testing for co re support structures are discussed in Subsections 3.9(N).2.3, 3.9(N).2.5 and 3.9(N).2.6. Stress limits and deformation criteria are given in Subsection 3.9(N).5.

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Page 60 3.9(N).3.1 Loading Combinations, Design Transients, and Stress Limits (for ASME Code Class 2 and 3 Components)

Design pressure, temperatures, 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. a. Design Loading Combinations The design loading combinations for ASME Code Class 2 and 3 components and supports are given in Table 3.9(N)-4. The design loading combinations are categorized with respect to Normal, Upset, Emergency, and Faulted Conditions. Stress limits for each of the loading combinations are component-oriented and are presented in Table 3.9(N)-5, Table 3.9(N)-6, Table 3.9(N)-7, Table 3.9(N)-8 and Table 3.9(N)-9 for tanks, inactive (operability is not relied upon to perform a safety function as defined for active in Section 3.9(N).2.2) pumps, active (operability is relied upon to perform a sa fety function as defined for active in Section 3.9(N).2.2) pumps, and valves, respectively. Active pumps and valves are discussed in Subsection 3.9(N).3.2. Design of component supports is discussed in Subsection 3.9(N).3.4. b. Design Stress Limits 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 Table 3.9(N)-5, Table 3.9(N)-6, Table 3.9(N)-7, Table 3.9(N)-8 and Table 3.9(N)-9. 3.9(N).3.2 Pump and Valve Operability Assurance

a. Pump and Valve Operability Program Mechanical equipment classified as safety-related must be capable of performing its function under postulated plant conditions. Equipment with faulted condition functional requirements includes active pumps and valves in fluid systems important to safety. Seismic analysis is presented in Section 3.7 and covers all safety-related mechanical equipment. A list of all active pumps supplied by Westinghouse is presented in Table 3.9(N)-10. Active valves supplied by Westinghouse are listed in Table 3.9(N)-11. The Westinghouse requirements necessary to implement the operability program described below are contained in the appropriate equipment specifications for active pumps and valves.

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Page 61 1. Pump Operability Program All active pumps are qualified for operability by first being subjected to 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 of the design pre ssure, (2) seal leakage tests at the same pressure used in the hydrostatic tests, and (3) 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. 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. After the pump is installed in the pl ant, it undergoes cold hydro tests, hot functional tests, and the required periodic in-service 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 during an SSE condition by assuring that the pump will continue operating and will not be damaged during the seismic event.

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Page 62 The pump manufacturer is required to show that the pump will operate normally when subjected to the maximum seismic accelerations and maximum faulted nozzle loads. It is required that tests 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 rigid. This frequency is considered sufficiently high to avoid problems with amplification between the component and structure for all seismic areas. A static shaft deflection analysis of the rotor is performed with the conserva tive SSE accelerations of 2.1g in two orthogonal horizontal directions a nd 2.1g vertical acceleration acting simultaneously. The deflections determined from the static shaft analysis are compared to the allowable rotor clearances. The nature of seismic disturbances dictates that the maximum contact (if it occurs) will be of short duration. To avoid damage dur ing the faulted plant condition, the stresses caused by the combination of normal operating loads, SSE, and dynamic system loads are limited as indicated in Table 3.9(N)-8. In addition, the pump casing stresses caused by the maximum seismic nozzle loads are limited to stresses outlined in Table 3.9(N)-8. The maximum seismic nozzle loads are also considered in an analysis of the pump supports to assure that a system misalignment cannot occur. Performing these analyses with the c onservative loads stated and with the restrictive stress limits of Table 3.9(N)-8 as allowables assures that critical parts of the pump will not be dama ged during the faulted condition and that, therefore, the reliability of the pump for post-faulted condition operation will not be impaired by the seismic event. The second criterion necessary to assure operability is that the pump will function throughout the SSE. The pump/motor combination is designed to rotate at a constant speed under all conditions unless the rotor becomes completely seized, i.e., with no rotation.

Typically, the rotor can be seized 5 full seconds before a circuit-breaker, to prevent damage to the motor, shuts down the pump. However, the high rotary inertia in the operating pump rotor, and the nature of the random, short duration loading characteristics of the seismic event, will prevent the rotor from losing its function. In actuality, the seismic loadings will cause only a slight increase, if any, in the torque (i.e., motor current) necessary to drive the pump at the constant design speed. Therefore, the pump will not shut down during the SSE and will operate at the design speed despite the SSE

loads.

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Page 63 To complete the seismic qualification procedures, the pump motor is independently qualified for operation during the maximum seismic event. Any auxiliary equipment which is identified to be vital to the operation of the pump or pump motor, and which is not qualified for operation during the pump analysis or motor qualifications, is also separately qualified for

operation at the accelerations it would see at its mounting. The pump motor and vital auxiliary equipment is qualified by meeting the requirements of IEEE Standard 344-1975, with the additional requirements and justifications outlined in Subsection 3.9(N).3.2b. 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 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 fau lted 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.

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Page 64 2. Valve Operability Program Active valves are subjected to a series of tests prior to service and during the plant life. Prior to installation, the following tests are performed: shell hydrostatic test to ASME Section III requirements, backseat and main seat leakage tests, disc hydrostatic test, a nd operational tests to verify that the valve will open and close. Qualification of motor operators for environmental conditions is discussed in Section 3.11(N). Cold hydro tests, hot functional qualification tests, periodic in-service inspections, and periodic in-service operations in accordance with the IST program, or

other Station procedures or Technical Specification surveillances 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. The maximum stress limits used for active Class 2 and 3 valves are shown in Table 3.9(N)-9. On active valves, an analysis of the extended structure is performed for static equivalent seismic loads applied at the center of gravity of the extended structure. In addition to these tests and anal yses, representative valves of each design type that were originally designa ted as active valves were 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 manne r which conservatively represents typical valve installations. The valv e included the operator pilot solenoid valves, and limit switches when such are normally attached to the valve in service. The faulted condition nozzle loads were limited so 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: (a) Active valves are designed to ha ve a first natural frequency which is greater than 33 Hz. This may be shown by test or analysis. For valves with a fundamental frequency below 33 Hz, the valve flexibility shall be accounted for in the subsystem analysis.

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Page 65 (b) The actuator and yoke of the valve system was statically deflected an amount equal to the deflecti on caused by the faulted condition accelerations applied at the cente r of gravity of the extended structure alone in the direction of the weakest axis of the yoke.

The design pressure of the valve was applied to the valve during the static deflection tests. (c) The valve was cycled while in the deflected position. The time required to open or close the valve in the deflected position was compared to similar design taken in the undeflected condition to

evaluate the significance of any change. (d) Motor operators, external limit sw itches, and pilot solenoid valves necessary for operation are qua lified in accordance with IEEE Standard 344-1975, with the additional requirements and

justifications as supplied in Subsection 3.9(N).3.2b. The accelerations which are used for th e static valve qualification shall be equivalent, as justified by analysis, to the simultaneous application of at

least 2.1g in two orthogonal horizontal di rections and at least 2.1g in the

vertical direction. The piping is designed to maintain the operator accelerations to these levels. The above testing program applied to valves with extended structures. The testing was conducted on a representative number of originally-designated active valves. Valves from each of the primary

safety-related design types were tested. Valve sizes which cover the range

of sizes in service were qualified by th e 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, are considered

separately.

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Page 66 Check valves are characteristically simple in design and their operation will not be affected by seismic accelerations or the maximum applied nozzle loads. The check valve design is compact and there are no extended structures or masses whose motion c ould cause distortions which could

restrict operation of the valve. The nozzle loads due to maximum 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 methods, the ability of the valve to operate is assured by the design features. The valve

will also undergo the following: (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 ar e qualified by the following procedures: (1) stress and deformation analyses of critical items which may affect

operability for faulted condition loads, (2) in shop hydrostatic and seat

leakage tests, and (3) periodic in situ valve inspection. In addition to these tests, a static load equivalent to that applied by the faulted condition is

applied at the top of the bonnet and the pressure is increased until the valve mechanism actuates. Succe ssful actuation within the design requirements of the valve assures its overpressurization safety capabilities during a seismic event. Using these methods, active valves are qualified for operability during a faulted event. These methods outlined above conservatively simulate the seismic event and assure that the active valves will perform their safety-related function when necessary. b. Pump Motor and Valve Operator Qualification Active pump motors (including vital pump appurtenances) and active valve motor operators are seismically qualified in accordance with IEEE Standard 344-1975.

The seismic qualification program for this electrical equipment is further described in Section 3.10(N) and the Equipment Qualification Data Packages

referenced therein. 3.9(N).3.3 Mounting of Pressure Relief Devices Refer to Subsection 3.9(B).3.3.

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Page 67 3.9(N).3.4 Component Supports (ASME Code Class 2 and 3)

See Subsection 3.9(N).1 for ASME Code Class 1 component supports. a. Component Supports for Components Procured After July 1, 1974 Class 2 and 3 component supports are designed and analyzed for Design, Normal, Upset, Emergency, and Faulted conditions to the rules and requirements of Subsection NF of Section III of the ASME B&PV Code (1974 Edition). The design analyses or test methods and associated stress or load allowable limits that are used in the evaluation of linear s upports for Faulted conditions are those defined in Appendix F of the ASME Code. b. Component Supports for Components Procured Prior to July 1, 1974 Class 2 and 3 supports are designed as follows: 1. Standard Component Supports (a) Normal - The allowable stresses or load ratings of MSS-SP-58 are used. (b) Upset - For upset conditions, the allowable stresses or load ratings are 20 percent higher than those specified for normal conditions. (c) Emergency - For emergency conditions, the allowable stresses or load ratings are 80 percent higher than those specified for normal

conditions. Supports (rod hangers and struts) are checked for elastic stability when applicable, and maximum compressive load

does not exceed critical buckli ng load as specified by the applicable codes and design standards. (d) Faulted - The allowable stresses or load ratings of MSS-SP-58 are based on a factor of safety which is greater than or equal to four, i.e., the allowable stress is less than or equal to one-fourth the minimum tensile stress of the material. The allowable stresses for faulted conditions are thus less than or equal to 0.6 times the minimum tensile stress of the material, i.e., 2.4 times one-fourth the minimum tensile stress of the material equals 0.6 times the minimum tensile stress. This low allowable stress (associated

factor of safety equals 1.67) is adequate to assure that active components are properly supported for faulted conditions.

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Page 68 2. Linear Type Supports (a) Normal - The allowable stresses of AISC-69 Part 1 are employed for normal condition limits. (b) Upset - Stress limits for upset conditions are 33 percent higher than those specified for normal conditions. This is consistent with paragraph 1.5.6 of AISC-69 Part 1 which permits one-third increase in allowable stresses for wind or seismic loads. (c) Emergency - Not applicable. (d) Faulted - Stress limits for faulted condition are the same as for the upset condition. 3. Plate and Shell Type Supports (a) Normal - Normal condition limits are those specified in ASME Section VIII Division 1 or AISC-69 Part 1. (b) Upset - Stress limits for upset conditions are 33 percent higher than those specified for normal conditions. This is consistent with paragraph 1.5.6 of AISC Part 1 which permits one-third increase in allowable stresses for wind or seismic loads. (c) Emergency - Not applicable. (d) Faulted - Stress limits for faulted condition are the same as for the upset condition. 3.9(N).4 Control Rod Drive System (CRDS) 3.9(N).4.1 Descriptive Information of CRDS

a. Control Rod Drive Mechanisms Control Rod Drive Mechanisms (CRDMs) are located on the head of the reactor vessel. They are coupled to Rod Cluster Control Assemblies which contain neutron absorber material over the entire length of the control rods. The control rod drive mechanism is shown in Figure 3.9(N)-3 and schematically in

Figure 3.9(N)-4.

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Page 69 The primary function of the CRDM is to insert or withdraw the Rod Cluster Control Assemblies within the core to control average core temperature and to shut down the reactor. The CRDM is a magnetically operated jack, which is an arrangement of three electromagnets energized in a controlled se quence by a power cycler, to insert or withdraw Rod Cluster Control Assemblies in the reactor core in discrete steps.

Rapid insertion of the Rod Cluster Control Assemblies occurs when electrical power is interrupted. The CRDM consists of four separate subassemblies. They are the pressure vessel, coil stack assembly, latch assembly, and the drive rod assembly. 1. Pressure Vessel The pressure vessel includes a la tch housing and a rod travel housing which are connected by a threaded, seal-welded maintenance joint which facilitates replacement of the latch assembly. The threaded connection forms the pressure boundary. The canopy seals are not pressure retaining and are designed to provide a means to control leaks through the threads.

If leakage develops through a canopy seal weld, either a weld repair or a canopy seal clamp assembly can be used to repair or prevent the leak. The closure at the top of the rod tr avel housing is a threaded plug with a canopy seal weld for pressure integrity. The latch housing is the lower portion of the vessel and contains the latch assembly. The rod travel housing is the upper portion of the vessel and

provides space for the drive rod during its upward movement as the control rods are withdrawn from the core. 2. Coil Stack Assembly The coil stack assembly includes the coil housings, an electrical conduit and connector, and three operating coils:

(1) the stationary gripper coil, (2) the movable gripper coil, and (3) the lift coil. The coil stack assembly is a separate unit, which is installed on the drive mechanism by sliding it over the outside of the latch housing. It rests on the base of the latch housing without mechanical attachment.

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Page 70 Energizing the operating coils causes movement of the pole pieces and latches in the latch assembly. 3. Latch Assembly The latch assembly includes the guide tube, stationary pole pieces, movable pole pieces, and two sets of latches: (1) the movable gripper latches and (2) the stationary gripper latches. The latches engage grooves in the drive rod assembly. The movable gripper latches are moved up or down in inch steps by the lift pole to raise or lower the drive rod. The stationa ry gripper latches hold the drive rod assembly while the movable gripper latches are repositioned for the next

inch step. 4. Drive Rod Assembly The drive rod assembly includes a flexible coupling, a drive rod, a disconnect device, a disconnect rod, and a locking button. The drive rod has inch grooves which receive the latches during holding or moving of the drive rod. The flexib le coupling is attached to the drive rod and provides the means for coup ling to the Rod Cluster Control Assembly. The disconnect device, disconnect rod, and locking button provide positive locking of the coupling to the Rod Cluster Control Assembly and permit remote uncoupling of the drive rod from the Rod Cluster Control Assembly. The control rod drive mechanism is a trip design. Tripping can occur during any part of the power cycler se quencing, if electrical power to the

coils is interrupted. The control rod drive mechanism is threaded and seal-welded on an adaptor housing on top of the reactor vessel, and is coupled to the Rod Cluster Control Assembly directly below. The mechanism is capable of raising or lowering a 360 pound load, (which includes the drive rod weight) at a rate of 45 inches/minute. Withdrawal of the Rod Cluster Control Assembly is accomplished by magnetic forces, while insertion is

by gravity.

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Page 71 The mechanism internals are designed to operate in 650F reactor coolant.

The pressure vessel is designed to contain reactor coolant at 650 F and 2500 . The three operating coils are designed to operate at 392F, with forced air cooling required to maintain that temperature as a normal maximum condition. The control rod drive mechanism, shown schematically in Figure 3.9(N)-4, withdraws and inserts a Rod Cluster Control Assembly as shaped electrical pulses are receive d by the operating coils. An ON or OFF sequence, repeated by silicon-controlled rectifiers in the power programmer, causes either withdrawal or insertion of the control rod. Position of the control rod is measured by 42 discrete coils mounted on the position indicator assembly surrounding th e rod travel housing. Each coil magnetically senses the entry and presence of the top of the ferromagnetic drive rod assembly as it moves through the coil center line. During plant operation the stationary gripper coil of the drive mechanism holds the Rod Cluster Control Assembly in a static position until a stepping sequence is initiated, at which time the movable gripper coil and lift coil are energized sequentially. b. Rod Cluster Control Assembly Withdrawal The Rod Cluster Control Assembly is withdrawn by repetition of the following sequence of events (refer to Figure 3.9(N)-4). 1. Movable Gripper Coil (B) - ON The latch-locking plunger raises and swings the movable gripper latches into the drive rod assembly groove. A 0.047 inch axial clearance exists between the latch teeth and the drive rod. 2. Stationary Gripper Coil (A) - OFF The force of gravity, acting upon the drive rod assembly and attached control rod, causes the stationary gripper latches and plunger to move downward 0.047 inch until the load of the drive rod assembly and attached control rod is transferred to the movable gripper latches. The plunger continues to move downward and

swings the stationary gripper latches out of the drive rod assembly groove.

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Page 72 3. Lift Coil (C) - ON The 5/8 inch gap between the movable gripper pole and the life pole closes and the drive rod assembly raises one step length (5/8 inch). 4. Stationary Gripper Coil (A) - ON The plunger raises and closes the gap below the stationary gripper pole. The three links, pinned to the plunger, swing the stationary gripper latches into a drive rod assembly groove. The latches contact the drive rod assembly and lift it (and the attached control rod) 0.047 inch. The 0.047 inch vertical drive rod assembly movement transfers the drive rod assembly load from the movable gripper latches to the stationary gripper latches. 5. Movable Gripper Coil (B) - OFF The latch locking plunger separates from the movable gripper pole under the force of a spring and grav ity. Three links, pinned to the plunger, swing the three movable gripper latches out of the drive rod assembly groove. 6. Lift Coil (C) - OFF The gap between the movable gripper pole and lift pole opens. The movable gripper latches drop 5/8 inch to a position adjacent to a drive rod assembly groove. 7. Repeat Step 1 The sequence described above (Items 1 through 6) is termed as one step or one cycle. The Rod Cluster Control Assembly moves 5/8 inch for each step or cycle. The sequence is repeated at a rate of up to 72 steps per minute, and the drive rod assembly (which has a 5/8 inch groove) is raised 72 grooves per minute. The Rod Cluster Control Assembly is t hus withdrawn at a rate up to 45 inches per minute.

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Page 73 c. Rod Cluster Control Assembly Insertion The sequence for Rod Cluster Control Assembly insertion is similar to that for control rod withdrawal, except the timing of lift coil (C) ON and OFF is changed to permit lowering the control assembly. 1. Lift Coil (C) - ON The 5/8 inch gap between the movable gripper and lift pole closes. The movable gripper latches are raised to a position adjacent to a drive rod assembly groove. 2. Movable Gripper Coil (B) - ON The latch-locking plunger raises and swings the movable gripper latches into a drive rod assembly groove. A 0.047 inch axial

clearance exists between the latch teeth and the drive rod assembly. 3. Stationary Gripper Coil (A) - OFF The force of gravity, acting upon the drive rod assembly and attached Rod Cluster Control Assembly, causes the stationary gripper latches and plunger to m ove downward 0.047 inch until the load of the drive rod assembly and attached Rod Cluster Control Assembly is transferred to the movable gripper latches. The plunger continues to move downward and swings the stationary gripper latches out of the drive rod assembly groove. 4. Lift Coil (C) - OFF The force of gravity and spring force separate the movable gripper pole from the lift pole and the drive rod assembly and attached rod cluster control drop down 5/8 inch.

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Page 74 5. Stationary Gripper (A) - ON The plunger raises and closes the gap below the stationary gripper pole. The three links, pinned to the plunger, swing the three stationary gripper latches into a drive rod assembly groove. The latches contact the drive rod assembly and lift it (and the attached control rod) 0.047 inch. The 0.047 inch vertical drive rod assembly movement transfers the drive rod assembly load from the movable gripper latches to the stationary gripper latches. 6. Movable Gripper Coil (B) - OFF The latch locking plunger separates from the movable gripper pole under the force of a spring and grav ity. Three links, pinned to the plunger, swing the three movable gripper latches out of the drive rod assembly groove. 7. Repeat Step 1 The sequence is repeated, as for Rod Cluster Control Assembly withdrawal, up to 72 times per mi nute which gives an insertion rate of 45 inches per minute. d. Holding and Tripping of the Control Rods During most of the plant operating time, the control rod drive mechanisms hold the Rod Cluster Control Assemblies withdrawn from the core in a static position. In the holding mode, only one coil, the stationary gripper coil (A), is energized on each mechanism. The drive rod assembly and attached Rod Cluster Control Assemblies hang suspended from the

stationary gripper latches. If power to the stationary gripper coil is cut off, the combined weight of the drive rod assembly and the Rod Cluster Control Assembly plus the stationary gripper return spring will push aside the latches and release the drive rod. The control rod falls by grav ity into the core. The trip occurs as the magnetic field, holding the st ationary gripper plunger against the stationary gripper fixed pole, collaps es and the stationary gripper plunger is forced down by the stationary gripper return spring and weight acting

upon the latches. After the Rod Cluster Control Assembly is released by the mechanism, it falls freely until the control rods enter the dashpot section of the thimble tubes in the fuel assembly.

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Page 75 3.9(N).4.2 Applicable CRDS Design Specifications For those components in the Control Rod Drive System comprising portions of the reactor coolant pressure boundary, conformance with the General Design Criteria and 10 CFR 50, Section 50.55a, is discussed in Sections 3.1 and 5.2. Conformance with pertaining regulatory guides is discussed in Sections 4.5 and 1.8 and in Subsection 5.2.3. a. Design Bases Bases for temperature, stress on structural members, and material compatibility are imposed on the design of the reactivity control components. b. Design Stresses The CRDS is designed to withstand stresses originating from various operating conditions, as summarized in Table 3.9(N)-1. Allowable Stresses: For normal operating conditions,Section III of the ASME Code is used. All pressure boundary components are analyzed as Class I components. c. Dynamic Analysis The cyclic stresses due to dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total stresses on the CRDS. d. Control Rod Drive Mechanisms The Control Rod Drive Mechanism (CRDM) pressure housings are Class I components designed to meet the stress requirements for normal operating conditions of Section III of the ASME Boiler and Pressure Vessel Code. Both static and alternating stress intensities are considered. The stresses originating from the required design transients are included in the analysis. A dynamic seismic analysis is required on the CRDMs when a seismic disturbance has been postulated, to confirm the ability of the pressure housing to meet ASME Code,Section III, allowable stre sses and to evaluate the effect of the seismic event on the drop time.

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Page 76 The control rod drive mechanisms (CRD Ms) are evaluated for the effects of postulated reactor vessel inlet branch line nozzle and outlet branch line nozzle breaks. A time-history analysis of the CRDMs is performed for the vessel motion discussed in Subsection 3.9(N).1.5. A model of the CRDMs is formulated with gaps at the upper CRDM supports modeled as nonlinear elements. The CRDMs are represented by beam elements with lumped masses. The translation and rotation of the vessel head is applied to this model. The resulting loads and stresses are compared to allowables to verify the adequacy of the system. e. Control Rod Drive Mechanism Operational Requirements The basic operational requirements for the CRDMs are: 1. 5/8 inch step 2. 147 inch travel

3. 360 pound maximum load
4. Step in or out at 45 inches/minute (72 steps/minute)
5. Electrical power interruption shall initiate release of drive rod assembly
6. Trip delay time of less than 150 milliseconds - free fall of drive rod assembly shall begin less than 150 milliseconds after power interruption, no matter what holding or stepping acti on is being executed with any load and coolant temperature of 100 F to 550 F 7. 40-year design life with normal refurbishment. 3.9(N).4.3 Design Loads, Stress Limits and Allowable Deformations
a. Pressure Vessel The pressure retaining components are analyzed for loads corresponding to normal, upset, emergency and faulted conditions. The analysis performed depends on the mode of operation under consideration. The scope of the analysis requires many different techniques and methods, both static and dynamic.

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Page 77 Some of the loads that are considered on each component where applicable are as follows: 1. Control rod trip (equivalent static load)

2. Differential pressure
3. Spring preloads
4. Coolant flow forces (static)
5. Temperature gradients
6. Differences in thermal expansion (a) Due to temperature differences (b) Due to expansion of different materials 7. Interference between components
8. Vibration (mechanically or hydraulically induced)
9. All operational transients listed in Table 3.9(N)-1
10. Pump overspeed
11. Seismic loads (Operating Basi s Earthquake and Safe Shutdown Earthquake) 12. Blowdown forces (due to cold and hot leg branch line breaks).

The main objective of the analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the functioning of the components.

The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consid eration of fatigue characteristics of the materials. Standard methods of strength of materials are used to establish the stresses and deflections of these components. The dynamic

behavior of the reactivity control components has been studied using experimental test data and experience from operating reactors.

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Page 78 b. Drive Rod Assembly All postulated failures of the drive rod assemblies either by fracture or uncoupling lead to a reduction in reactivity. If the drive rod assembly fractures at any elevation, that portion remaining coupled falls with, and is guided by the Rod Cluster Control Assembly. This always results in reactivity decrease for the control rods. c. Latch Assembly and Coil Stack Assembly

1. Results of Dimensional and Tolerance Analysis With respect to the CRDM system as a whole, critical clearances are present in the following areas: Latch assembly (diametral clearances)

Latch arm-drive rod clearances Coil stack assembly-thermal clearances Coil fit in coil housing. The following paragraphs define clear ances that are designed to provide reliable operation in the control rod drive mechanism in these four critical areas. These clearances have been proven by life tests and actual field performance at operating plants.

(a) Latch Assembly - Thermal Clearances The magnetic jack has several clearances where parts made of Type 410 stainless steel fit over parts made from Type 304 stainless steel. Differential thermal expansion is therefore important. Minimum clearan ces of these parts at 68 F is 0.011 inches. At the maximum design temperature of 650 F, minimum clearance is 0.0045 inches; at the maximum expected operating temperature of 550F, minimum clearance is 0.0057 inches.

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Page 79 (b) Latch Arm - Drive Rod Clearances The CRDM incorporates a load transfer action. The movable or stationary gripper latches are not under load during engagement, as previously explained, due to load transfer action. Figure 3.9(N)-5 shows latch clear ance variation with the drive rod as a result of minimum and maximum temperatures.

Figure 3.9(N)-6 shows clearance variations over the design temperature range. (c) Coil Stack Assembly - Thermal Clearances The assembly clearances of the coil stack assembly over the latch housing were selected so that the assembly could be removed under all anticipated conditions of thermal expansion. At 70F, the inside diameter of the coil stack is 7.308/7.298 inches, and the outside diameter of the latch housing is 7.260/7.270 inches. Thermal expansion of the mechanism, due to operating temperature of the CRDM, results in the minimum inside diameter of the coil stack being 7.310 inches at 222F and the maximum latch housing outside diameter being 7.302 inches at 532 F. Under the extreme tolerance cond itions listed above, it is necessary to allow time for a 70F coil housing to heat during a replacement operation. To verify the acceptability of the above tolerances, four coil stack assemblies were removed from four hot control rod drive mechanisms, mounted on 11.035 inch centers on a 550 F test loop, allowed to cool, and then placed without incident. (d) Coil Fit in Coil Housing CRDM and coil housing clearances ar e selected so that coil heat up results in a close to tight fit. This is done to facilitate thermal transfer and coil cooling in a hot control rod drive mechanism.

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Page 80 3.9(N).4.4 Evaluation of Control Rod Drive Mechanisms and Supports The Control Rod Drive Mechanisms (CRDMs) and CRDM support structure are evaluated for the loading combinations outlined in Table 3.9(N)-2. A detailed finite element model of the CRDMs and CRDM supports is constructed using the WECAN computer program with beam, pipe, and spring elements. For the LOCA analysis, nonlinearities in the structure are represented. These include RPI plate impact, tie rods, and lifting leg clevis/RPV head interface. The time-history motion of the reactor vessel head, obtained from the RPV analysis is input to the dynamic model. Maximum forces and moments in the CRDMs and support structure are then determined. For the seismic analysis, the structural model is linearized and the floor response spect ra corresponding to the CRDM tie rod elevation is applied to determine the maximum forces and moments in the structure. The bending moments calculated for the CRDMs for the various loading conditions are compared with maximum allowable moments determined from a detailed finite element stress evaluation of the CRDMs. Adequacy of the CRDM support structure is verified by comparing the calculated stresses to the criteria given in ASME III, Subsection NF. 3.9(N).4.5 CRDS Performance Assurance Program

a. Evaluation of Material's Adequacy The ability of the pressure housing components to perform throughout the design lifetime, as defined in the equipment specification, is confirmed by the stress analysis report required by the ASME Code,Section III. Internal components subjected to wear will withstand a minimum of 3,000,000 steps without refurbishment, as confirmed by life tests (Reference 12). Latch assembly inspection is recommended after 2.5x10 6 steps have been accumulated on a single control rod drive mechanism. To confirm the mechanical adequacy of the fuel assembly, the control rod drive mechanism and Rod Cluster Control Assembly, functional test programs have been conducted on a full-scale 12-foot c ontrol rod. The 12-foot prototype assembly was tested under simulated conditions of reactor temperature, pressure, and flow for approximately 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />. The prototype mechanism accumulated about 3,000,000 steps and 600 trips. At the end of the test, the control rod drive mechanism was still operating satisfactorily. A correlation was developed to predict the amplitude of flow-excited vibr ation of individual fuel rods and fuel assemblies. Inspection of the drive line components did not reveal significant fretting.

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Page 81 These tests include verification that the trip time achieved by the CRDM meets the design requirement of 2.2 seconds from start of Rod Cluster Control Assembly motion to dashpot entry. This trip time requirement will be confirmed for each control rod drive mechanism prior to initial reactor operation and at periodic intervals after initial reactor operation, as required by the proposed

Technical Specifications. These tests have been reported in Reference 12. In addition, dynamic testing programs have been conducted by Westinghouse and Westinghouse Licensees to demonstrate that control rod scram time is not adversely affected by postulated seismic events. Acceptable scram performance is assured by also including the effects of the allowable displacements of the driveline components in the evaluation of the test results. There are no significant differences be tween the prototype control rod drive mechanisms and the production units. Design materials, tolerances and fabrication techniques are the same (see Section 4.5). It is expected that all control rod drive mechanisms will meet specified operating requirements for the duration of plant life, with normal refurbishment. Latch assembly inspection is recommended after 12.5x10 6 steps have been accumulated on a single CRDM. If a Rod Cluster Control Assembly cannot be moved by its mechanism, adjustments in the boron concentration ensure that adequate shutdown margin would be achieved following a trip. Thus, inability to move one Rod Cluster Control Assembly can be tolerated. More than one inoperable Rod Cluster Control Assembly could be tolerated, but would impose additional demands on the plant operator. Therefore, the num ber of inoperable Rod Cluster Control Assemblies has been limited to one as discussed in the proposed Technical Specifications. In order to demonstrate proper operation of the control rod drive mechanism, and to ensure acceptable core power dist ributions during Rod Cluster Control Assembly partial-movement, checks are performed on the Rod Cluster Control Assemblies (refer to Technical Specifications

). In addition, periodic drop tests of the Rod Cluster Control Assemblies are performed at each refueling shutdown, to demonstrate continued ability to meet trip time requirements. During these tests, the acceptable drop time of each assembly is not greater than 2.4 seconds, at full flow and operating temperature, from the beginning of motion to dashpot entry.

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Page 82 Actual experience in operating Wes tinghouse plants indicates excellent performance of control rod drive mechanisms. All units are production tested prior to shipment, to confirm ab ility of the control rod mechanism to meet design specification operation requirements. Each production CRDM undergoes a production test as listed below:

Test Acceptance Criteria Cold (ambient) hydrostatic ASME Section III Confirm step length and load transfer (stationary gripper Step Length 5/8 +/-0.015 inches axial movement to movable gripper or movable

gripper to stationary gripper)

Load Transfer 0.047 inches nominal axial movement Cold (ambient) performance

test at design load -

Operating Speed 45 inches/minute 5 full travel excursions Trip Delay Free fall of drive rod to begin within 150 milliseconds S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 83 3.9(N).5 Reactor Pressure Vessel Internals 3.9(N).5.1 Design Arrangements The reactor vessel internals are described as follows:

The components of the reactor internals are divide d into three parts consisting of the lower core support structure (including the entire core barrel and neutron shield pad assembly), the upper core support structure and the incore instrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and control rod drive mechanisms, direct coolant flow past the fuel elements, direct coolant flow to the pressure vessel head, provide gamma and neutron shielding, and provide guides for the incore instrumentation. The coolant flows from the vessel inlet nozzles down the annulus between the core barrel and the vessel wall and then into a plenum at the bottom of the vessel. It then reverses and flows up through the core support and through the lower core plate. The lower core plate is sized to provide the desired inlet flow distribution to the core. After passing through the core, the coolant enters the region of the upper support structure and then flows radially to the core barrel outlet nozzles and directly through the vessel outlet nozzles. A small portion of the coolant flow s between the baffle plates and the core barrel to provide additional cooling of the barrel. Similarly, a small amount of the entering flow is directed into the vessel head plenum and exits through the vessel outlet nozzles. a. Lower Core Support Structures The major containment and support member of the reactor internals is the lower core support structure, shown in Figure 3.9(N)-7. This support structure assembly consists of the core barrel, the core baffle, the lower core plate and support columns, the neutron shield pads, and the core support which is welded to the core barrel. All the major material for this structure is Type 304 stainless steel.

The lower core support structure is supported at its upper flange from a ledge in the reactor vessel head flange, and its lowe r end is restrained in its transverse movement by a radial support system attached to the vessel wall. Within the core

barrel are an axial baffle and a lower core plate, both of which are attached to the core barrel wall and form the enclosure periphery of the assembled core. The lower core support structure and principally the core barrel serve to provide

passageways and control for the coolant flow. The lower core plate is positioned at the bottom level of the core below the baffle plates and provides support and orientation for the fuel assemblies.

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Page 84 The lower core plate is a member thr ough which the necessary flow distribution holes for each fuel assembly are machined. Fuel assembly locating pins (two for each assembly) are also inserted into this plate. Columns are placed between this

plate and the core support of the core barrel to provide stiffness and to transmit the core load to the core support. Ad equate coolant distribution is obtained through the use of the lower core plate and core support. The neutron shield pad assembly consists of four pads that are bolted and pinned to the outside of the core barrel. These pads are constructed of Type 304 stainless steel, and are approximately 48 inches wide by 148 inches long by 2.8 inches thick. The pads are located azimuthally to provide the required degree of vessel protection. Specimen guides in which material surveillance samples can be inserted and irradiated during reactor ope ration are attached to the pads. The samples are held in the guides by a preloaded spring device at the top and bottom to prevent sample movement. Additional details of the neutron shield pads and irradiation specimen holders are given in Reference 13. Vertically downward loads from weight, fuel assembly preload, control rod dynamic loading, hydraulic loads and ear thquake acceleration are carried by the

lower core plate partially into the lowe r core plate support flange on the core

barrel shell and partially through the lower support columns to the core support and, thence, through the core barrel shell to the core barrel flange supported by the vessel head flange. Transverse loads from earthquake acceleration, coolant

cross flow, and vibration are carried by the core barrel shell and distributed between the lower radial support to the vessel wall, and to the vessel flange.

Transverse loads of the fuel assemblies are transmitted to the core barrel shell by direct connection of the lower core plat e to the barrel wall and by upper core plate alignment pins which are welded into the core barrel. The Main Radial Support System of the lower end of the core barrel is accomplished by "key" and "keyway" joints to the reactor vessel wall. At equally spaced points around the circumference, an In conel clevis block is welded to the vessel inner diameter. Another Inconel in sert block is bolted to each of these blocks and has a "keyway" geometry. O pposite each of these is a "key" which is attached to the internals. At assembly, as the internals are lowered into the vessel, the keys engage the keyways in the axial direction. With this design, the internals

are provided with a support at the furthest extremity, and may be viewed as a beam supported at the top and bottom.

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Page 85 Radial and axial expansions of the core barrel are accommodated but transverse movement of the core barrel is restricted by this design. With this system, cyclic stresses in the internal structures are within ASME Code,Section III, limits. In the event of an abnormal downward vertical displacement of the internals following a hypothetical failure, energy-absorbing devices limit the displacement after contacting the vessel bottom head. Th e load is then transferred through the energy-absorbing devices of the internals to the vessel. The energy absorbers, cylindrical in shape, are contoured on their bottom surface to the reactor vessel bottom head geometry. Assuming a downward vertical displacement the potential energy of the system is absorbed mostly by the strain energy of the energy-absorbing devices. b. Upper Core Support Assembly The upper core support assembly, shown in Figure 3.9(N)-8 and Figure 3.9(N)-9, consists of the top support plate assemb ly and the upper core plate, between which are contained support columns and guide tube assemblies. The support columns establish the spacing between the top support plate assembly and the upper core plate, and are fastened at top and bottom to these plates. The support columns transmit the mechanical loadings between the two plates and serve the supplementary function of supporting thermocouple guide tubes. The guide tube assemblies sheath and guide the control rod drive shafts and control rods. They

are fastened to the top support plate, and are restrained by pins in the upper core plate for proper orientation and support.

Additional guidance for the control rod drive shafts is provided by the upper guide tube which is attached to the upper

support plate and guide tube.

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Page 86 The upper core support assembly is positioned in its proper orientation with respect to the lower support structure by fl at-sided pins pressed into the core barrel which, in turn, engage in slots in the upper core plate. At an elevation in the core barrel where the upper core plate is positioned, the flat-sided pins are located at angular positions of 90 degrees from each other. Four slots are milled into the core plate at the same positi ons. As the upper support structure is lowered into the main internals, the slots in the plate engage the flat-sided pins in the axial direction. Lateral displacem ent of the plate and of the upper support assembly is restricted by this design. Fuel assembly locating pins protrude from the bottom of the upper core plate and engage the fuel assemblies as the upper assembly is lowered into place. Proper alignment of the lower core support structure, the upper core support assembly, the fuel assemblies and control rods are thereby assured by this system of locating pins and guidance arrangement. The upper core support assembly is restrained from any axial movements by a large circumferential spring which rests between the upper barrel flange and the upper core support assembly and is compressed by the reactor vessel head flange. Vertical loads from weight, earthquake acceleration, hydraulic loads and fuel assembly preload are transmitted thr ough the upper core plate via the support columns to the top support plate assembly and then the reactor vessel head.

Transverse loads from coolant cross fl ow, earthquake acceleration, and possible vibrations are distributed by the support columns to the top support plate and upper core plate. The top support plate is particularly stiff to minimize deflection. c. Incore Instrumentation Support Structures The incore instrumentation support structure consists of a guide tubing system to convey and support flux thimbles penetrating the vessel through the bottom. Each detector assembly includes five fixed gamma/neutron flux detectors, a calibration tube for a movable neutron flux detector, and a thermocouple. (Figure 7.7-9 shows the basic Flux-Mapping System.) Replacement detector assemblies do not have a complete calibration tube and cannot be traversed by a moveable detector.

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Page 87 The guide tubes extend from the seal ta ble down through the concrete shield area and terminate in socket welds at the reactor vessel bottom head penetrations. The guide tube bend radius is 12 feet. The detector assemblies extend through the guide tubing and vessel penetrations, through hollow passages in the lower internals and finally through instrumentation support tubes in the fuel assemblies. The detector assemblies remain in pl ace during operation but are pulled back approximately 13 feet at the seal tabl e during refueling to avoid interference within the core. The detector assemblies ar e closed at the leading ends and sealed against the guide tubes at the seal table. Mechanical seals between the retractable detector assemblies and guide tubes are provided at the seal table. During normal operation, the retractable detector assemblies are stationary and move only during refueling or for maintenance, at which time a space of approximately 15 feet above the seal table is cleared for the retraction operation. The incore instrumentation support structure is designed for adequate support of instrumentation during reactor operation and is rugged enough to resist damage or distortion under the conditions imposed by handling during the refueling sequence. These are the only conditions which affect the incore instrumentation support structure. 3.9(N).5.2 Design Loading Conditions The design loading conditions that provide the basi s for the design of the reactor internals are: a. Fuel assembly weight b. Fuel assembly spring forces

c. Internals weight
d. Control rod trip (equivalent static load)
e. Differential pressure
f. Spring preloads
g. Coolant flow forces (static)
h. Temperature gradients S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 88 i. Differences in thermal expansion due to: 1. Temperature differences 2. Expansion of different materials j. Interference between components

k. Vibration (mechanically or hydraulically induced)
l. One or more loops out of service
m. All operational transients listed in Table 3.9(N)-1
n. Pump overspeed
o. Seismic loads (Operating Basis Earthquake and Safe Shutdown Earthquake)
p. Blowdown forces (due to cold and hot leg branch line breaks). The main objective of the design analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the functioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but to also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials. Both low and high cycle fatigue stresses are considered when the allowable amplitude of oscillation is established. Dynamic analysis on the reactor internals is provided in Subsection 3.9(N).2.

As part of the evaluation of design loading c onditions, extensive testing and inspections are performed from the initial selection of raw materials up to and including component installation and plant operation. Among these tests and inspections, are those performed during component fabrication, plant construction, start up and checkout, and during plant operation. 3.9(N).5.3 Design Loading Categories The combination of design loadings fit into either the normal, upset, emergency or faulted conditions, as defined in the ASME Code,Section III. However, it should be noted that by

contract the reactor internals for the Seabrook plant preceed the applicability of Subsection NG of the ASME Code. Therefore, these internals are not "Code Stamped" and no specific stress report is required. Nevertheless, these reactor internals are designed to meet the intent of Subsection NG of the ASME Code.

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Page 89 Loads and deflections imposed on components due to shock and vibration are determined analytically and experimentally in both scaled models and operating reactors. The cyclic stresses due to these dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total stresses of the internals.

The reactor internals are designed to withstand stresses originating from various operating conditions, as summarized in Table 3.9(N)-1.

The scope of the stress analysis problem is very large requiring many different techniques and methods, both static and dynamic. The analysis performed depends on the mode of operation under consideration. 3.9(N).5.4 Design Bases The design bases for the mechanical design of the reactor vessel internals components are as follows: a. The reactor internals in conjunction with the fuel assemblies direct reactor coolant through the core to achieve acceptable fl ow distribution and to restrict bypass flow so that the heat transfer performance requirements are met for all modes of operation. In addition, required cooling for the pressure vessel head is provided so that the temperature differences be tween the vessel flange and head do not result in leakage from the flange during reactor operation. b. In addition to neutron shielding provi ded by the reactor coolant, a separate neutron pad assembly is provided to limit the exposure of the pressure vessel, in order to maintain the required ductility of the material for all modes of operation. c. Provisions exist for installing incore instrumentation useful for the plant operation and vessel material test specimens required for a pressure vessel irradiation surveillance program. d. The core internals are designed to withstand mechanical loads arising from the Operating Basis Earthquake, Safe Shutdown Earthquake and pipe ruptures, and meet the requirement of Item e. below. e. The reactor has mechanical provisions which are sufficient to adequately support the core and internals, and to assure that the core is intact with acceptable heat transfer geometry following transients arising from abnormal operating conditions.

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Page 90 f. Following the design basis accident, the pl ant shall be capable of being shut down and cooled in an orderly fashion so that fuel cladding temperature is kept within specified limits. This implies that the deformation of certain critical reactor internals must be kept sufficiently small to allow core cooling. The functional limitations for the core structures during the design basis accident are shown in Table 3.9(N)-12. To ensure no column loading of rod cluster control guide tubes, the upper core plate deflection is limited to not exceed the value shown in Table 3.9(N)-12. Details of the dynamic analyses, input forcing f unctions, and response loadings are presented in Subsection 3.9(N).2.

The basis for the design stress and deflection criteria is identified below: a. Allowable Stresses For normal operating conditions,Section III of the ASME Code is used as a basis for evaluating acceptability of calculated st resses. Both static and alternating stress intensities are considered. It should be noted, that the allowable stresses in Section III of the ASME Code are based on unirradiated material properties.

In view of the fact that irradiation increases the strength of the Type 304 st ainless steel used for the internals, although decreasing its elongation, it is c onsidered that use of the allowable stresses in Section III is appropriate a nd conservative for irradiated internal structures. The allowable stress limits during the de sign basis accident used for the reactor internals are based on the draft of the 1971 edition of the ASME Code for Core Support Structures, Subsection NG, and the Criteria for Faulted Conditions. b. Allowable Deflections For normal operating conditions, downward ve rtical deflection of the lower core support plate is negligible. For the loss-of-coolant accident plus the Safe Shutdown Earthquake condition, the deflection criteria of critical internal structures are the limiting values given in Table 3.9(N)-12. The corresponding no-loss-of-function limits are included in Table 3.9(N)-12 for comparison purposes with the allowed criteria.

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Page 91 The criteria for the core drop accident is based upon analyses which have to determine the total downward displacement of the internal structures following a hypothesized core drop resulting from loss of the normal core barrel supports.

The initial clearance between the secondary core support structures and the reactor vessel lower head in the hot condition is approximately 1/2 inch. An additional displacement of approximately 3/4 inch would occur due to strain of the energy absorbing devices of the seconda ry core support; thus the total drop distance is about 11/4 inches, which is insufficient to permit the tips of the Rod Cluster Control Assembly to come out of the guide thimble in the fuel assemblies. Specifically, the secondary core support is a device which will never be used, except during a hypothetical accident of the core support (core barrel, barrel flange, etc.). There are 4 supports in each reactor. This device limits the fall of the core and absorbs much of the ener gy of the fall which otherwise would be imparted to the vessel. The energy of the fall is calculated assuming a complete and instantaneous failure of the primary core support and is absorbed during the plastic deformation of the controlled volume of stainless, loaded in tension. The maximum deformation of this austenitic stainless piece is limited to approximately 15 percent, after which a positive stop is provided to ensure support. 3.9(N).6 In-Service Testing of Pumps and Valves Refer to Subsection 3.9(B).6.

3.9(N).7 References

1. "Documentation of Selected Westinghouse Structural Analysis Computer Codes," WCAP-8252, Revision 1, May 1977. 2. "Benchmark Problem Solutions Employed for Verification of the WECAN Computer Program," WCAP-8929, June 1977. 3. "Sample Analysis of a Class 1 Nuclear Piping System," prepared by ASME Working Group on Piping, ASME Publication, 1972. 4. Witt, F. J., Bamford, W. H., Esselman, T. C., "Integrity of Primary Piping Systems of Westinghouse Nuclear Power Plants during Postulated Seismic Events," WCAP-9283, March 1978.

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Page 92 5. Bogard, W. T., Esselman, T. C., "Combination of Safe Shutdown Earthquake and Loss-of-Coolant A ccident Responses for Faulted Condition Evaluation of Nuclear Power Plants," WCAP-9279, March 1978. 6. "Prediction of the Flow-Induced Vi bration of Reactor Internals by Scale Model Tests," WCAP-8303-P-A (Proprietary) and WCAP-8317A (Nonproprietary), July 1975. 7. Bloyd, C. N. and Singleton, N.

R., "UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations," WCAP-8516-P (Proprietary) and WCAP-8517 (Nonproprietary), April

1975. 8. Bloyd, C. N., Ciaramitaro, W. and Singleton, N. R., "Verification of Neutron Pad and 17x17 Guide Tube De signs by Preoperational Tests on the Trojan 1 Power Plant," WCAP-8766 (Proprietary) and WCAP-8780 (Nonproprietary), May 1976. 9. Takeuchi, K., et al., "MULTIFLEX A FORTRAN-IV Computer Program for Analyzing Thermal Hydraulic-Structure System Dynamics," WCAP-8708-P-A, Westinghouse Proprietary Class 2 / WCAP-8709 NES Class 3 (Non-Proprietary), September 1977. 10. Takeuchi, K., et al., "MULTIFLEX 3.0 A FORTRAN IV Computer Program for Analyzing Thermal Hydraulic-Structural System Dynamics Advanced Beam Model," WCAP-9735, Revision 2, Westinghouse Proprietary Class 2/ WCAP-9736, Revi sion 1, Non-Proprietary, February 1998. 11. Bohm, G. J. and LaFaille, J. P

., "Reactor Internals Response Under a Blowdown Accident," Procedures, First Intl. Conf. on Structural Mech. in Reactor Technology, Berlin, September 20-24, 1971. 12. Cooper, F. W. Jr., "17x17 Drive Line Components Tests - Phase IB, II, III, D-Loop Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Nonproprietary), December 1974. 13. Kraus, S., "Neutron Shielding Pads," WCAP-7870, May 1972. 14. Final Rule Modifying 10 CFR 50 Appendix A, GDC-4 dated October 27, 1987 [52 FR 41288].

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Mechanical Systems and Components Revision 13 Section 3.9(N)

Page 93 15. Letter, T. H. Essig (USNRC) to Lou Liberatori (WOG), "Safety Evaluation of Topical Report WCAP-15029, 'Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions', (TAC NO. MA1152)," November 10, 1998 (Enclosure 1 - Safety Evaluation Report). 16. Schwirian, R. E., et al., "W estinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barre l Bolting Distributions Under Load Conditions," WCAP-15029-P-A, We stinghouse Proprietary Class 2/WCAP-15030-NP-A, Revision 0, non-proprietary, January 1999.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical E q ui pment Revision 8 Section 3.10(B) Pa g e 1 3.10(B) SEISMIC QUALIFICATION OF CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT This section discusses the seismic qualification criteria, methods and procedures employed by the A-E for the qualification of seismic Category I instrumentation and electrical equipment within his scope of responsibility. It also covers the methods of analysis or testing of the supports for the electrical equipment and instrumentation. Seismic Category I instrumentation and electrical equipment are listed in Table 3.2-1 and Table 3.2-2. The corresponding discussion for the electrical equipment, instrumentati on and supports provided by the NSSS supplier is found in Section 3.10(N).

3.10(B).1 Seismic Qualification Criteria The criteria employed for seismic qualification of seismic Category I (Class 1E) electrical equipment and instrumentation, other than NSSS-related equipment, follow the guidelines recommended in IEEE Std. 344-1975 and Regulatory Guide 1.100. The seismic Category I instrumentation and electrical equipment are designed to withstand, without loss of function or structural integrity, the combined effects of all normal operating loads and the seismic loads of the Sa fe Shutdown Earthquake (SSE) or the Operating Basis Earthquake (OBE), as defined in Subsection 3.7(B).1. The seismic Category I instrumentation and electrical equipment are seismically qualified by using either analytical methods or testing, or by a combination of both, as follows:

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical E q ui pment Revision 8 Section 3.10(B) Pa g e 2 a. The equipment which does not undergo a ch ange of state is qualified by analysis or testing to establish structural adequacy. This includes cable tray systems, cable conduit systems and bus duct systems. When analytical methods are used to qualify the seismic Category I equipment or component, stresses resulting from the seismic acceleration effects are combined with stresses due to normal operating loads. Both horizontal and vertical seismic loads are assumed to occur simultaneously in the most unfavorable combinations, and the two horizontal and one vertical seismic load-induced stress components are combined by the square-root-of-the-sum-of-the-squares me thod. The stress levels due to the combined normal design loads and seismic loads are maintained within the stress levels set forth in appropriate design sta ndards and codes. If there are no code requirements in the design of the equipment or portions thereof, then the stress level under the above combined loading, including the normal design loads plus SSE seismic load condition, is limited to 90 percent of the minimum yield strength for the material. In addition, the equipment or component design is reviewed to assure that any resulting defl ections or distortions do not prevent the proper functioning of the equipment, nor endanger adjacent equipment or components. The criteria for selecting a static load analysis or dynamic modal analysis are decided by the complexity of the equipment, as delineated in IEEE Std. 344-1975. When testing methods are used to qualify cable tray systems, the testing performance will meet the intent of IEEE Std. 344-1975. b. The equipment or components which must be capable of undergoing a change of state are qualified by a combination of seismic and functional testing both during and after an earthquake of magnitude up to and including the SSE. The equipment or components are tested for ab ility to retain structural integrity, and no malfunction is permitted where such a malfunction could jeopardize the

capability to safely shut down the reactor and/or mitigate offsite exposures. When testing is used to qualify seismic Category I equipment or components, testing is performed in accordance with IEEE Std. 344-1975. c. Seismic Category I equipment supports, including cabinets, panels, consoles and instrumentation racks are qualified by either analysis or testing using appropriate

horizontal and vertical floor response sp ectra at the building and elevation at which they are installed. When analysis is used to qualify the equipment supports, the stress criteria are in accordance with AISC Manual of Steel Construction.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical E q ui pment Revision 8 Section 3.10(B) Pa g e 3 d. Seismic qualification tests are conducted for battery prototypes with cables (or equivalent batteries with cables) in accordance with the requirements of IEEE 323-1974 and IEEE 344-1975. The battery racks are qualified by appropriate seismic analyses which include the battery masses. The Class 1E transformers are seismically qualified by test using an appropriate test apparatus

which includes the supporting structures or cabinets and all relevant

appurtenances including cooling accessories. This effort is included in the

Class 1E Unit Substation Qualification as noted in Table 3.10(B)-1.

3.10(B).2 Methods and Procedures for Qualifyi ng Instrumentation and Electrical Equipment Seismic Category I instrumentation and electrical equipment (other than NSSS) were qualified either by analysis, by testing, or by a combinati on of testing and analysis as indicated in Table 3.10(B)-1, to confirm their functional opera bility during and after an earthquake up to and including the SSE. All specifications for seismic Category I instrumentation and electrical equipment included the seismic design criteria are discussed in Subsection 3.10(B).1 as a part of the design condition. Tests or analyses are performed in accord ance with the criteria and the intent of IEEE Std. 344-1975 and Regulatory Guide 1.100. Amplified floor response spectra applicable to the particular equipment locations were included as a part of the specification. The seismic intensities indicated in these spectra were used by the manufacturer for seismic qualification of the equipment and capability documentation.

Certified test report and/or analytical calculations were obtained from each vendor to confirm that the purchased seismic Category I instrumentation and electrical equipment will perform its function when subjected to the stipulated seismic loading conditions of the SSE. All test reports and calculations were certified by a registered professional engineer, skilled in the applicable specialty, and by a responsible officer of the manufacturer or vendor. Equipment anchor loadings and details, such as size and spacing of anchor bolts or welds, were obtained from the equipment manufacturers for use in designing foundations or supporting floors for compatibility with the seismic anchor loading of the equipment.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical E q ui pment Revision 8 Section 3.10(B) Pa g e 4 3.10(B).3 Methods and Procedures of Analysis or Testing of Supports of Electrical Equipment and Instrumentation 3.10(B).3.1 Electrical Equipment and Instrument Supports The qualification of supports for main control boards, cabinets, panels and instrument racks, as well as supports for electrical equipment such as battery racks, was accomplished using one of the methods discussed in Subsection 3.10(B).1. The methods used in evaluating the supports were tested under simulated conditions and analytical approaches. Analytical methods were employed to anchor the supports. Amplified floor response spectra for the locations where the equipment is mounted are provided to the equipment supplier who is responsible for qualifying the equipment. Supports for instruments and electric equipment are attached by bolting or welding to anchor plates fabricated of ASTM A36 steel, either embedded in the concrete with stud anchors or surface-mounted to the concrete using bolt anchors. In either case, they were designed to prevent uplift or overturning effects due to seismic forces.

3.10(B).3.2 Cable Tray Supports Raceway systems when used to carry safety-rela ted circuit cables, are designed or tested to withstand the seismic forces which would be expe rienced during an SSE due to the weight of the cables, raceways and supports. Cable tray load-deflection curves were used to formulate a simplified analytical model of the tray which was then coupled to the analytical model of the supports. Load deflection parameters may be established by testing to formulate the analytical models of the cable tray supports. The response spectra method was used to analyze the overall analytical model and to design the support structures, while complying with the tray support system functional requirement. The Cable Tray Support System was analyzed for dead load combined with the OBE loads, with the stress criteria based on the allowable stresses of the AISC Specification on structural steel for buildings and the engineering information for strut members published by Unistrut, Powerstrut or Superstrut. For dead loads combined with SSE loads, the stresses are limited to 90 percent of yield stress for the material involved. For the seismic loads, the actual natural frequency response of each support system was calculated and the appropriate seismic acceleration factor was selected from the amplified floor response spectra (see Section 3.7).

The cable tray supports consist of structural shapes and strut members. The structural steel portion of the supports is fabricated from ASTM A36 steel. Each support is connected to either the structural steel framing, the concrete floor slab or the concrete wall of the building structures.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical E q ui pment Revision 8 Section 3.10(B) Pa g e 5 3.10(B).4 Operating License Review Results obtained from the seismic qualification tests and/or analyses performed for the various categories of equipment listed in Table 3.10(B)-1 are documented in the reference documents (see contract no.) listed in the same table. These documents are maintained in the Information Management System files, and in UE&C Foreign Print files as required by ANSI N45.2.9.

United Engineers and Constructors has reviewed these documents as part of the routine vendor document review process to assure that the seismic qualification methods are acceptable, and the results thereof sufficiently demonstrate the qualification of the equipment.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical Equipment Revision 8 Section 3.10(N) Page 1 3.10(N) SEISMIC QUALIFICATION OF CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT This section presents information to demonstrate that instrumentation and electrical equipment classified as seismic Category I is capable of performing designated safety-related functions in the event of a safe shutdown earthquake (SSE). The information presented includes identification of the seismic Category I instrumentation and electrical equipment that are within the scope of the Westinghouse Nuclear Steam Supply System (NSSS), the seismic qualification criteria employed and, for each item of equipmen t, the designated safety-related functional requirements, definition of the applicable seismic environment and documentation of the qualification process employed to demonstrate the required seismic capability.

3.10(N).1 Seismic Qualification Criteria 3.10(N).1.1 Qualification Standards The methods of meeting the general requirements for seismic qualification of Category I instrumentation and electrical equipment, as described by General Design Criteria (GDC) 1, 2 and 23, are described in Section 3.1. The general methods of implementing the requirements of Appendix B to 10 CFR Part 50 are described in Chapter 17. The Commission's recommendations concerning the methods to be employed for seismic qualification of electrical equipment are contained in Regulatory Guide 1.100, which endorses IEEE 344-1975. Westinghouse meets this standard, as modified by Regulatory Guide 1.100, by either type test, analysis, or an appropriate combination of these methods. Westinghouse will meet this commitment employing the methodology de scribed in the final staff approved version of Reference 1.

3.10(N).1.2 Performance Requirements for Seismic Qualification Reference 2 contains an equipment qualification data package (EQDP) for every item of instrumentation and electrical equipment classified as seismic Category I within the Westinghouse NSSS scope of supply. Table 3.10(N)-1 identifies the seismic Category I equipment supplied by Westinghouse for this appli cation and references the applicable EQDP contained in Supplement 1. Each EQDP in Supplement 1 contains a section entitled "Performance Specification." This specification establishes the safety-related functional requirements of the equipment to be demonstrated during and after a seismic event. The required response spectrum (RRS) employed by Westinghouse for generic seismic qualification

is also identified in the specification, as applicable. The spectra employed have been selected to envelope the plant specific spectra defined in Section 3.7.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical Equipment Revision 8 Section 3.10(N) Page 2 3.10(N).1.3 Acceptance Criteria Seismic qualification must demonstrate that seismic Category I instrumentation and electrical equipment is capable of performing designated safety-related functions during and after an earthquake of magnitude up to and including th e Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) without the initiation of undesired spurious actuation which might result in consequences adverse to safety. The qualification must also demonstrate the structural integrity of mechanical supports and structures at the OBE level. Some permanent mechanical deformation of supports and structures is acceptabl e at the SSE level providing that the ability to perform the designated safety-related functions is not impaired.

3.10(N).2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation In accordance with IEEE 344-1975, seismic qualification of safety-related electrical equipment is demonstrated by either type testing, analysis or a combination of these methods. The choice of qualification method employed by Westinghouse for a particular item of equipment is based upon many factors including: practicability, complexity of equipment, economics, availability of previous seismic qualification to earlier standards, etc. The qualification method employed for a particular item of equipment is identified in the individual Equipment Qualification Data Packages (EQDPs) of Reference 2.

3.10(N).2.1 Seismic Qualification by Type Test From 1969 to mid-1974 Westinghouse seismic test procedures employed single axis sine beat inputs in accordance with IEEE 344-1971 to seismically qualify equipment. The input form selected by Westinghouse was chosen following an investigation of building responses to seismic events as reported in Reference 3. In addition, Westinghouse has conducted seismic retesting of certain items of equipment as part of the Supplemental Qualification Program (Reference 4). This retesting was performed at the request of the NRC st aff on agreed selected items of equipment employing multi-frequency, multi-axis test inputs (Reference 5), to demonstrate the conservatism of the original sine beat test method with respect to the modified methods of testing for complex equipment recommended by IEEE 344-1975.

The original single axis sine beat testing and the additional retesting completed under the Supplemental Test Program has been the subject of generic review by the Staff. For equipment which has been previously qualified by the single axis sine beat method and included in the NRC seismic audit and, where required by the Staff, the Supplemental Qualification Program (Reference 4), no additional qualification testing is required to demonstrate acceptability to IEEE 344-1975 provided that:

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical Equipment Revision 8 Section 3.10(N) Page 3 a. The Westinghouse aging evaluation program for aging effects on complex electronic equipment located outside containment demonstrates there are no deleterious aging phenomena. In the event that the aging evaluation program identifies materials that are marginal, either the materials will be replaced or the projected qualified life will be adjusted. b. Any changes made to the equipment due to a. above or due to design modifications do not significantly affect the seismic characteristics of the equipment. c. The previously employed test inputs can be shown to be conservative with respect to applicable plant specific response spectra. This equipment is identified in Reference 1, Table 7.1 and the test results in the applicable EQDPs of Reference 2. For equipment tests after July 1974 (i.e., new designs, equipment not previously qualified or previously qualified equipment that does not meet a, b, and c above), seismic qualification by test is performed in accordance with IEEE 344-1975. Where testing is used, multi-frequency multi-axis inputs are developed by the general procedures outlined in Reference 5. The test results contained in the individual EQDPs of Reference 2 demonstrate that the measured Test Response Spectrum envelopes the applicable Required Response Spectrum (RRS) defined for generic testing, as specified in Section I of the EODP (Reference 2). Qualification for plant-specific use is established by verification that the generic RRS specified by Westinghouse envelopes the applicable plant specific response spectrum. Alternative test methods such as single frequency and single axis inputs are used in selected cases as permitted by IEEE 344-1975 and Regulatory Guide 1.100.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical Equipment Revision 8 Section 3.10(N) Page 4 3.10(N).2.2 Seismic Qualification by Analysis The structural integrity of safety-related motors (Table 3.10(N)-1, EQDP-AE-2 and 3) is demonstrated by a static seismic analysis in accordance with IEEE 344-1975, with justification.

Should analysis fail to show the resonant frequency to be significantly greater than 33 Hz, a test is performed to establish the motor resonant frequency. Motor operability during a seismic event is demonstrated by calculating critical deflections, loads and stresses under various combinations of seismic, gravitational and operational loads. The worst case (maximum) values calculated are tabulated against the allowable values. On combining these stresses, the most unfavorable possibilities are considered in the following areas: (1) maximum rotor deflection, (2) maximum shaft stresses, (3) maximum bearing load and shaft slope at the bearings, (4) maximum stresses in the stator core welds, (5) maximum stresses in the stator core to frame welds, (6) maximum stresses in the motor mounting bolts and (7) maximum stresses in the motor feet. The analytical models employed and the results of the analysis are described in Section 4 of the applicable EQDPs (Reference 2).

3.10(N).3 Method and Procedures for Qualifyi ng Supports of Electrical Equipment and Instrumentation Where supports for the electrical equipment and instrumentation are within the Westinghouse NSSS scope of supply, the seismic qualification te sts and/or analysis are conducted including the supplied supports. The EQDPs contained in Reference 2 identify the equipment mounting employed for qualification purposes and establish interface requirements for the equipment to ensure that subsequent in-plant installation does not prejudice the qualification established by Westinghouse.

3.10(N).4 Operating License Review The results of tests and analyses that ensure that the criteria established in Subsection 3.10(N).1 have been satisfied employing the qualification methods described in Subsections 3.10(N).2 and 3.10(N).3 are included in the individual EQDPs contained in Reference 2 as they become available.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Seismic Qualification of Category I Instrumentation and Electrical Equipment Revision 8 Section 3.10(N) Page 5 3.10(N).5 References

1. Butterworth, G. and Miller, R.B., "Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment," WCAP-8587, Revision 2, February 1979. 2. "Equipment Qualification Data Packages," Supplement I to WCAP-8587, November 1978. 3. Morrone, A., "Seismic Vibration Testing with Sine Beats," WCAP-7558, October 1971. 4. NS-CE-692, Letter dated July 10, 1975 from C. Eicheldinger (Westinghouse) to D.B. Vasello (NRC). 5. Jarecki, S.J., "General Method of Developing Multi-Frequency Biaxial Test Inputs for Bistables," WCAP-8624 (Proprietary) September 1975 and WCAP-8695 (Nonproprietary) August 1975.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Environmental Design of Mechanical and Electrical Equipment Revision 10 Section 3.11 Page 1 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT Safety-related equipment must be capable of maintaining functional operability under conditions postulated to occur during its installed life. This requirement is embodied in 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," 10 CFR 50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," and 10 CFR 50 Appendix A, "General Design Criteria for Nuclear Power Plants," Criteria 1, 2, 4, 23 and 50. The NRC has used a variety of methods to ensure that these general requirements are met for electric equipment important to safety. For nuclear plants after 1971, qualification was judged on the basis of IEEE Std 323-1971, "Trial Use Standard - General Guide for Qualifying Class I Electric Equipment for Nuclear Power Generating Stations." For plants whose construction permit Safety Evaluation Reports (CP-SER) were issued after July 1, 1974, the Commission has used Regulatory Guide 1.89 which endorses IEEE Std 323-1974, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations," subject to supplementary

provisions. The NRC subsequently issued more definitive criteria in NUREG-0588, "Interim Staff Position on Environmental Qualification (EQ) of Safety-Related Electrical Equipment," which contains two sets of criteria: a. Category I, for plants whose CP-SER was issued after July 1, 1974, incorporates and supplements IEEE Std 323-1974 b. Category II, for plants whose CP SER was issued before July 1, 1974, incorporates and supplements IEEE Std 323-1971.

Because the Seabrook CP-SER was issued August 14, 1974, NUREG-0588, Category I Criteria are applicable.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Environmental Design of Mechanical and Electrical Equipment Revision 10 Section 3.11 Page 2 Environmental qualification of safety-related equipment located in a mild environment is ensured by conformance to the general quality and surveillance requirements identified in 10 CFR, Part 50, Appendix B, and Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," as implemented in Chapter 17. A separate report, "Environmental Qualification of Electrical Equipment Important to Safety" (E.Q. Report), has been developed which provides the detailed information required to certify qualification of electrical components located in a potential harsh environment (Reference 1). The report describes the scope of the Seabrook E.Q. Program, including the station operations and maintenance E.Q. program. The E.Q. program development, methodology and the equipment qualification status is also discussed in the reports. The Mechanical Equipment Qualification status was submitted January 1984 (Reference 2). A revised final Mechanical Equipment Qualification (MEQ) report was submitted in April 1986.

The NSSS-supplied equipment and BOP-supplied equipment have been evaluated for the Seabrook plant specific environments by the same E.Q. programs. The E.Q. Reports and MEQ Reports describe the methodology and results for all plant equipment. 3.11.1 Determination of Equipment to be Environmentally Qualified The list of systems and equipment was established by considering those components required to safely shut down and mitigate the following accidents: a. Loss-of-Coolant Accident (LOCA) b. Main Steam Line Break (MSLB)

c. Fuel Handling Accident (FHA)
d. High Energy Line Break (HELB)
e. Moderate Energy Line Break (MELB).

Specifically included according to NRC guidance are those systems required to achieve or support the following: a. Emergency Reactor Shutdown

b. Containment Isolation
c. Reactor Core Cooling S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Environmental Design of Mechanical and Electrical Equipment Revision 10 Section 3.11 Page 3 d. Containment Heat Removal e. Core Residual Heat Removal
f. Prevention of Radioactive Releases to the Environment. Section 2.0 of the E.Q. Report describes the scope and criteria for listing Electrical Equipment on the Harsh Environment Equipment List. Appendix A of the E.Q. Report contains the actual list of equipment. The list of mechanical equipment required to operate during and following an accident is in Appendix A of a report entitled "Environmental Qualification of Mechanical Equipment," submitted via Reference 8. 3.11.2 Establishment of Environmental Service Conditions Environmental conditions have been determined for normal and accident conditions. The specific normal, abnormal, accident and post-accident environments for Seabrook Station have been calculated for various environmental z ones of all safety-related buildings. The environmental parameters of interest are temperature, pressure, humidity, radiation, chemical spray and submergence. Section 3.0 of the E.Q. Report describes in detail the methodology and assumptions used in calculating the various building zone environments. The maximum, normal, and minimum environmental parameters are provided in Updated FSAR Figure 3.11-1 (Figure 3-1, "Service Environment Chart," of the E.Q. Report). Table 3-2 of the E.Q. Report is a summary of flood levels in the safety-related buildings which could result from postulated accidents.

The accident radiation doses have been ev aluated up to an average core burnup of 45,000 MWD/MTU and an analyzed core power of 3659 MWT. The plant HVAC systems were designed to maintain the temperature and humidity within the normal/accident limits defined in the Service Environment Chart. The HVAC system design is described in detail in Section 9.4. Most safety-related equipment necessary to perform required safe shutdown and accident mitigation functions is located in areas where the environment is controlled by redundant safety-related ventilation systems. This assures that the environment and therefore the safety-related equipment located in these areas will not be adversely affected by a ventilation equipment failure. The environmental parameters for areas where redundant, safety-related ventilation systems are not provided reflect an assumed loss of ventilation in the Service Environment Chart. This assures that safety-related equipment located in these areas is qualified to function properly in an environment resulting from a loss of ventilation.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Environmental Design of Mechanical and Electrical Equipment Revision 10 Section 3.11 Page 4 3.11.2.1 Main Steam Line Break (MSLB) Eff ect on Environmental Qualification of Equipment Environmental conditions due to certain size MSLBs postulated both inside and outside of containment cause superheated temperatures in the immediate environmental zone. In June 1984, Westinghouse issued a letter/notice of a possible unreviewed safety question (Reference 3). Later in December 1984, NRC issued IE Information Notice No. 84-90 regarding the same subject for plant review and applicability (Reference 4).

Seabrook Station has reviewed these conditions and has determined that the environmental parameters postulated for MSLB inside containment are not exceeded. The result of Westinghouse superheat sensitivity analyses presented in WCAP-8822, Supplement 2, shows that the addition of superheat has a negligible impact on containment pressure and temperature for atmospheric and sub-atmospheric containment designs. This result is due primarily to the following characteristics: a. Containment peak pressure and temperature are predominately affected by the total mass and energy release to the containment b. The addition of superheat has no impact on steamline break mass releases and minimal impact on energy release. Westinghouse has further stated that the above characteristics are independent of the steam generator type. Although the results and conclusions of WCAP-8822, Supplement 2, are based on a Model 51 steam generator, they are applicable to the Seabrook Model F steam generator. The review of MSLB outside of containment for Seabrook indicates the main steam and feedwater pipe chases are the only affected building locations.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Environmental Design of Mechanical and Electrical Equipment Revision 10 Section 3.11 Page 5 A closer detailed review of the MS&FW pipe chase design at Seabrook has shown that Seabrook can achieve a safe shutdown under any postulated superheated temperature profile due to a MSLB. This is achieved principally by the sepa ration criteria conceptually designed into these building areas. Seabrook has two separated MS&FW pipe chase areas exiting the east and west sides of containment. Each pipe chase houses the feedwater and main steam piping for two of the four steam generators. The piping is desi gned under the concepts of "superpipe" (i.e., low stress allowables and upgraded ISI Program). Since the Westinghouse requirement is for a minimum two steam generator cooldown, Seabr ook could safely shutdown under a postulated MSLB in the MS&FW pipe chase designed with "superpipe," using the alternate pipe chase.

The safe shutdown is described in detail in the Seabrook Station Appendix R report (Reference 5). The MS&FW pipe chase houses the MS&FW containment isolation valves, main steam safety valves, atmospheric dump valves and MS supply valves to the emergency feed pump turbine. This equipment has been Environmentally Qualified to perform its design basis function during a postulated MSLB outside Containment. 3.11.3 Qualification Testing and Analysis of Electrical Equipment Electrical equipment was evaluated to ensure th at it will function as required when exposed to normal and postulated accident environments. All testing and analysis conformed to the requirements of IEEE 323-1974 and the guidance of NUREG-0588, Category I. The various testing parameters to demonstrate qualification of equipment include: functional criteria, test sequence, aging methodology, accident environment, margins, electrical connection interface. These parameters and their application to the Seabrook qualification testing program are described to a greater extent in Section 4.0 of the E.Q. Report. The NSSS-supplied components were tested by a generic equipment qualification program conducted by Westinghouse. This program was described by Westinghouse WCAP-8587, Revision 2, and has been reviewed and accepted by the NRC (Reference 6). The Westinghouse program resulted in generic environmental qualification data packages (EQDP) for NSSS electrical equipment. These have been issued as Supplement 1 to WCAP-8587 (Reference 7).

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Environmental Design of Mechanical and Electrical Equipment Revision 10 Section 3.11 Page 6 3.11.4 Methodology for Evaluating Environmen tal Qualification to Plant Service Conditions A comparison of environmental qualification test results to equipment service conditions in the Seabrook plant was completed for all equipment located in harsh environmental zones. This engineering analysis results in a documentation file, called an Environmental Qualification File (EQF), which certifies qualification to the plant environments. A description of the scope and content of a typical EQF is detailed in Section 4.2 of the E.Q. Report. Each Westinghouse generic EQDP is evaluated as any other vendor test report, and a corresponding Seabrook EQF is

established to verify site specific qualification. In addition, a verification program has been developed and completed to ensure that the vendor equipment tested was either identical or similar to the installed plant equipment.

The results of the Seabrook Qualification Analysis are described in Appendix B "Qualification Evaluation Worksheets" of the E.Q. Report. 3.11.5 Operational Phase and Maintenan ce of Environmental Qualification Program The purpose of the Seabrook Station Equipment Qualification Program is the preservation of the qualification of equipment that is important to safety. In order to accomplish the task, the station has developed Design Control, Procurement and Maintenance Programs. Each program has incorporated the requirements of environmen tal qualification according to the functional requirements of the program. The station programs and procedures are prepared to maintain proper design control for plant modifications, procurement of new equipment and spare parts. The station maintenance program is designed to provide preventative as well as corrective maintenance which is identified by field operational experience. Section 5.0 of the E.Q. Report provides a more detailed discussion of the operational and maintenance programs.

3.11.6 References

1. PSNH letter SBN-886, "Equipment Qualification of Electrical Equipment; SER Outstanding Issue No. 6," dated October 31, 1985, from John DeVincentis to George W. Knighton. 2. PSNH letter SBN-608, "Mechanical Equipment Qualification," dated January 5, 1984, from John DeVincentis to George W. Knighton.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS Environmental Design of Mechanical and Electrical Equipment Revision 10 Section 3.11 Page 7 3. Westinghouse letter NAH-2495, "Environmental Qualification of Equipment for High Energy Line Breaks Outside of Containment," dated June 11, 1984, from D.P. Adomaitis to R. J. DeLoach. 4. NRC IE Information Notice No. 84-90, "Main Steam Line Break Effect on Environmental Qualification of Equipment," dated December 7, 1984. 5. PSNH letter SBN-904, "Fire Protection of Safe Shutdown Capability," dated December 2, 1985, from John DeVincentis to George W. Knighton. 6. Butterworth, G. and Miller, R.B., "Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment," WCAP-8587, Revision 2, February 1979. 7. "Equipment Qualification Data Packages," Supplement 1 to WCAP-8587, November 15, 1978. 8. PSNH letter SBN-1005, "Mechanical Equipm ent Qualification," dated April 16, 1986, from John DeVincentis to Vincent N. Noonan. 9. PSNH letter SBN-988, "Environmental Qualif ication; Post Accident Operability Time," dated April 3, 1986, from John DeVincentis to Vincent S. Noonan. 10. PSNH SBN-998, "Response to Environmenta l Qualification Audit Observation," dated April 10, 1986, from John DeVincentis to Vincent S. Noonan. 11. PSNH letter SBN-1031, "Environmental Qualification, "Evaluation Worksheets," dated May 7, 1986, from John DeVincentis to Vincent S. Noonan.

S EABROOK S TATION U PDATED F INAL S AFETY A NALYSIS R EPORT C HAPTER 3 DESIGN OF STRUCTURES

, COMPONENTS

, EQUIPMENT AND SYSTEMS T ABLES S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.2-1 Revision:

Sheet: 13 1 of 6 TABLE 3.2-1 SEISMIC CATEGORY I STRUCTURES , SYSTEMS AND COMPONENTS A. SEISMIC CATEGORY I STRUCTURES SYSTEM AND COMPONENTS (See Note 4) UFSAR Reference Section 1. Containment Structure Cylinder Dome Base Mat Liner Plate 3.8.1 2. Containment Internal Structures, including Fill Mat and Emergency Sump and Debris Strainer, and Debris Interceptors 3.8.3 3. Other Seismic Category I Structures Containment Enclosure Building Containment Equipment Hatch Missile Shield Containment Enclosure Ventilation Area Control and Diesel Generator Building Control Room Makeup Air Intake Structures Emergency Feedwater Pump Building, including Electrical Cable Tunnels and Penetration Areas (Control Building to Containment)

Enclosure for Condensate Storage Tank Fuel Storage Building Main Steam and Feedwater Pipe Chase (East), including East Penetration Area Main Steam and Feedwater Pipe Chase (West), including Mechanical Penetration Area and Personnel Hatch Area Piping Tunnels Pre-Action Valve Building Primary Auxiliary Building, including Residual Heat Removal (RHR) Equipment Vault Safety-Related Electrical Duct Banks and Manholes Service Water Access Vault Service Water Cooling Tower, including Switchgear Rooms Service Water Pumphouse Tank Farm (Tunnels), including Dikes and Foundations for Refueling Water Storage Tank (RWST) and Reactor Makeup Water Storage Tank Waste Processing Building Circulating Water Pumphouse Concrete below El 21'-0 3.8.4 4. Foundations for Seismic Category I Structures 3.8.5 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.2-1 Revision:

Sheet: 13 2 of 6 B. SEISMIC CATEGORY I MECHANICAL EQUIPMENT AND COMPONENTS SYSTEM AND COMPONENTS UFSAR Reference Section 1. Containment Equipment Equipment Hatch Personnel Lock Mechanical Penetrations Electrical Penetrations 3.8.2 2. ASME Code Class 1, 2 and 3 Component Supports and Core Support Structures 3.9.3 3. Control Rod Drive System Control Rod Drive System CRDM Seismic Support Platform (see Note 1) CRDM Seismic Support Space Plates (see Note 1) CRDM Seismic Support Tie Rod Assemblies (see Note 1) 3.9.4 4. Reactor Pressure Vessel Internals 3.9.5 5. Reactor Vessel Reactor Vessel Head Studs, Nuts and Washers Reactor Vessel Shoes and Shims Irradiation Sample Holder (see Note 2) 5.3 6. Incore Instrumentation Seal Table Assembly Guide Tubing (see Note 3)) 7.7.1.9 7. New Fuel Storage Racks 9.1.1 8. Spent Fuel Storage Spent Fuel Pool Spent Fuel Storage Racks 9.1.2 9. Fuel Handling System Reactor Vessel Head and Upper Internals Lifting Device (portions that furnish support to CFDMs) Spent Fuel Handling Tool Fuel Transfer System Fuel Transfer Tube and Flange Fuel Transfer Tube Outer Sleeve Expansion Joints Cask Handling Crane Trolley Frame and Main Hoist Machinery required to hold the load. 9.1.4 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.2-1 Revision:

Sheet: 13 3 of 6 C. SEISMIC CATEGORY I FLUID SYSTEMS AND COMPONENTS SYSTEM AND COMPONENTS UFSAR Reference Section Seismic Category I Fluid Systems and Components are identified in Table 3.2-2 D. SEISMIC CATEGORY I ELECTRICAL SYSTEMS AND COMPONENTS UFSAR Reference Section 1. Instrumentation and Controls Reactor Trip Switchgear Process Control Equipment Cabinets Westinghouse Solid-State Protection System and Safeguard Test Cabinets Nuclear Instrumentation System Cabinets Excore Neutron Detectors (Power Range and Wide Range)

Pressure Transmitters Differential Pressure Transmitters Resistance Temperature Detectors Thermocouples Main Control Boards Remote Shutdown Panels Safety-Related Pilot Solenoid Valves Safety-Related Externally Mounted Limit Switches Instruments and Controls for use in Seismic Category I systems and components which are required to perform nuclear safety-related functions Instrumentation Supports, Fittings and Accessories Hydrogen Recombiner Package Airborne and Particulate Radioactivity Monitors (APRM) Online Purge Monitor Containment Atmosphere Monitor Control Room Air Intake Monitor Area Radiation Monitors Containment Area Monitor Refueling Canal Monitor (mounted on Manipulator Crane) Accident Monitoring Instrumentation (Design Category I) Seismic Monitoring Instrumentation Reactor Vessel Level Indication System High Energy Line Break Thermocouples Containment Building Level Transmitters Main Steam Isolation Valve Logic System Post Accident Sampling Isolation Valve Control Panel Remote Safe Shutdown Disabling Control Panel Radiation Monitoring Isolation Valves Control Panels Inadequate Core Cooling Monitor 7.0

6.2.5 12.3.4 12.3.4 7.5

4.4.6.5.7.5

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.2-1 Revision:

Sheet: 13 4 of 6 D. SEISMIC CATEGORY I ELECTRICAL SYSTEMS AND COMPONENTS (Continued) UFSAR Reference Section 2. Onsite Power Systems 4160V Switchgear (ESF Buses) 4160V Nonsegregated bus duct between ESF buses and diesel generators 4000V and 460V Motors (associated with ESF) Diesel Generators including those auxiliaries necessary for operation, e.g. governor, voltage regulator and excitation system. (See Note 6) Diesel Generators Control Panels 480V Motor Control Centers (associated with ESF) 480V Unit Substations (ESF buses) 4160V to 480V Transformers (associated with ESF) 120V Vital Panel Boards Containment Penetration Assemblies including primary and backup fault current protective devices. (See Note 7)

Power Cables (5 kV and 600V) Instrumentation and Control Cables associated with ESF, including underground cable systems, e.g., cables in duct banks and cable splices. Emergency Power Sequencing System Conduit and Cable Tray Raceway System (nuclear safety-related) (See Note 5) 125V Batteries and battery racks (nuclear safety-related) Battery Chargers (nuclear safety-related) 125V DC Switchgear (nuclear safety-related) 125V DC Panelboards (nuclear safety-related)

Inverters (vital instrument buses)

Electrical Supports, Fittings and Accessories (i.e. connectors, terminal blocks, etc.) nuclear safety-related) 8.3.1, 8.3.2

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.2-1 Revision:

Sheet: 13 5 of 6 E. SEISMIC CATEGORY I HVAC SYSTEMS AND COMPONENTS UFS AR Reference Section 1. Control Building Control Room Air Conditioning System Remote Air Intake Ventilation System Cable Spreading Room Ventilation System Battery Room Ventilation System Switchgear Room Ventilation System 2 Diesel Generator Building Diesel Generator Room Ventilation System

3. Fuel Storage Building Fuel Storage Building Ventilation System
4. Primary Auxiliary Building Equipment Vault Area Ventilation System Containment Enclosure Area Ventilation System
5. Service Water Pumphouse Service Water Pumphouse Building Ventilation System
6. Miscellaneous Areas Main Steam and Feed Water Pipe Chases (East and West)

Ventilation System Mechanical Penetration Area Ventilation System Emergency Feed Water Pumphouse Building Ventilation System Cooling Tower Ventilation System

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.2-1 Revision:

Sheet: 13 6 of 6 SEISMIC CATEGORY I STRUCTURES, SYSTEMS AND COMPONENTS NOTES

1. These items not required as mechanical suppor ts for CRDM housings, but are required to ensure functioning of the control rods. 2. Any reactor vessel internal, the single failure of which could cause release of a mechanical piece having potential for direct damage (as to the vessel cladding) or flow blockage, shall be classified to a minimum of Safety Class 2 (see Subsection 3.2.2.1 for definition), seismic Category I. 3. Failure could cause a loss-of-coolant acciden t, but less than a C ondition III loss of coolant. 4. All seismic Category I structures are founde d either on sound bedrock or on engineered backfill extending to sound bedroc
k. The type of engineered backfill used beneath the foundations of all seismic Category I structures was fill concrete, except for safety-related electrical duct banks, electrical manholes and service water pipes which were founded on offsite borrow or tunnel cu ttings, as shown in Table 2.5-20. 5. The conduit and cable tray raceway systems including their supports, when used to carry safety-related circuit cables and other raceway installations whose failure during a seismic event could damage other safety-related systems or components are seismically qualified as assemblies. The items that make up the supports and cable trays are treated as nonsafety-related structural members, but are purchased as a component with specific performance requirements. The manufactur er provides substantiating test data and calculations, as well as a certificate of compliance to his manufacturing standards. The supports are assembled, installed and inspected in accordance with the applicable criteria of 10 CFR 50 Appendix B. The cable trays are designed, installed and inspected in accordance with Regulatory Position C.2 and C.4 of Regulatory Guide 1.29, Revision 3. Qualification of the conduit and cable tray raceways for the Class 1E safety related circuits have been confirmed by analysis and/or testing, both of these methods verify the adequacy of the system based on the proper ties of the raceways (including tray where applicable) and support components. 6. Components that are part of the diesel p ackage but not essential to the operation i.e., electric motors for auxiliary coolant pump and auxiliary lube oil pump etc., are not

included in this category. 7. See UFSAR Subsection 8.3.1.1.a4 for exception pertaining to the protection of the 13.8 kV containment electrical penetrations.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 1 of 42 TABLE 3.2-2 SEISMIC AND SAFETY CLASSIFICATIONS FLUID SYSTEMS AND COMPONENTS UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Reactor Coolant System Reactor Vessel Vessel and Closure Head 1 ASME III (7) 1 I CS W Full Length CRDM Housing 1 ASME III 1 I CS W CRDM Head Adapter Plugs 1 ASME III 1 I CS W 5.3 Incore Instrumentation Guide Tubes 1 ASME III 1 I CS W Reactor Coolant Pump Pump Casing 1 ASME III 1 I CS W Thermal Barrier 1 ASME III 1 I CS W Thermal Barrier Heat

Exchanger 1 ASME III 1 I CS W No. 1 Seal Housing 1 ASME III 1 I CS W No. 2 Seal Housing 2 ASME III 1 I CS W See Note 4. Pressure Retaining Bolting 1 ASME III 1 I CS W 5.4.1 Pump Flywheel 3 N/A N/A I CS W Steam Generator Tube (Reactor Coolant) Side 1 ASME III 1 I CS W

5.4.2 Shell

(Steam) Side 2 ASME III 1 I CS W See Note 4.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 2 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Reactor Coolant Piping Hot and Cold Leg Piping, Fittings and Fabrication 1 ASME III 1 I CS W For safety class for other piping and associated

valves in the Reactor Coolant System and other auxiliary systems, see

Note 3. Crossover Leg Piping, Fittings and Fabrication 1 ASME III 1 I CS W Surge Pipe, Fittings and Fabrication 1 ASME III 1 I CS W Reactor Coolant Thermowells 1 ASME III 1 I CS W 5.4.3 Reactor Coolant System

Vents 2 ASME III 1 I CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 3 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Residual Heat Removal

System Residual Heat Removal Pumps 2 ASME III 2 I PB W See Notes 1a, 1c, and 2. Residual Heat Exchanger Tube Side 2 ASME III 2 I PB W Shell Side 3 ASME III 3 I PB W See Note 1c. Piping and Valves Reactor Coolant Pressure Boundary 1 ASME III 1 I CS AE See Note 3. Sample/Drain Lines NNS ANSI B31.1 - - PB AE 5.4.7 Other 2 ASME III 2 I PB/CS AE See Note 3. 5.4.10 Pressurizer 1 ASME III 1 I CS W See Note 1b. Pressurizer Relief Discharge

System Pressurizer Relief Tank NNS ASME VIII - I CS W 5.4.11 Piping (Downstream of Safety

Valves) NNS ANSI B31.1 - I CS AE Valves of Safety Class 1 to

Safety Class 2 Interface 1 ASME III 1 I CS W 5.4.12 Valves of Safety Class 2 2 ASME III 2 I PB/CS W

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 4 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Safety and Relief Valves Pressurizer Safety Valves 1 ASME III 1 I CS W Pressurizer Power-Operated Relief Valves 1 ASME III 1 I CS W Block Valves for Power-

Operated Relief Valves 1 ASME III 1 I CS W 5.4.13 Piping Upstream of Pressurizer Safety and Relief

Valves 1 ASME III 1 I CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 5 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Engineered Safety Features Containment Heat Removal

Systems Containment Spray System Containment Spray Pumps 2 ASME III 2 I PB AE Spray Additive Tank 3 ASME III 3 I WB AE Containment Spray Heat

Exchangers Tube Side 2 ASME III 2 I PB AE Shell Side 3 ASME III 3 I PB AE Refueling Water Storage

Tank 2 ASME III 2 I WB AE Spray Headers and Nozzles 2 ASME III 2 I CS AE Piping and Valves 2 ASME III 2 I CS/PB/WP AE See Note 3.

6.2.2 3 ASME III 3 I CS/PB AE Containment Isolation

System 6.2.4 Penetration Piping and Isolation Valves included as part of the Containment Isolation System 2 ASME III MC I CS/CE AE (22)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 6 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Combustible Gas Control in

Containment Piping and

Valves(pressurized) 2 ASME III 2 I CS AE Containment Recirculating Filter System Fans 3 AMCA - I CS AE Filter Unit NNS ANSI N509/

N510 - - CS AE Ductwork and Dampers 3 SMACNA, ANSI N509/N510 - I CS AE Containment Purge System All pressurized components upstream of emergency exhaust filters (flow meter, throttle valve, piping) 2 ASME III 2 I CS/CE AE Containment Atmosphere Sample System Supply and Exhaust Piping 2 ASME III 2 I CS/CE AE

6.2.5 Supply

and Exhaust Valves 2 ASME III 2 I CS/CE AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 7 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Emergency Core Cooling

System Accumulator 2 ASME III 2 I PB W Safety Injection Pump 2 ASME III 2 I PB W See Notes 1a, 1c & 2. Piping Reactor Coolant Pressure Boundary 1 ASME III 1 I CS/PB W See Note 3. Drain, vent and test lines beyond the first pressure boundary isolation valve. NNS ANSI B31.1 - - PB AE

6.3 Valves

2 ASME III and ANSI B16.5 2 I PB/CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 8 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Engineered Safety Feature

Filter Systems Containment Enclosure Emergency Cleanup System Fans 2 AMCA - I CE AE Filters (HEPA & Charcoal) 2 ANSI N509/

N510 - I CE AE Ductwork and Dampers 2 ANSI N509/

N510 - I CE AE Fuel Storage Building Emergency Cleanup System Fans 3 AMCA - I FB AE Filters (Charcoal & HEPA) 3 ANSI N509/

N510 - I FB AE Ductwork & Dampers 3 ANSI N509/

N510 - I FB AE Control Room Emergency Filtration Subsystem Fans 3 AMCA - I CD AE/PSNH Filters (HEPA & Charcoal) 3 ANSI N509/

N510 - I CD AE/PSNH

6.5.1 Ductwork

and Dampers 3 ANSI N509/

N510 - I CD AE/PSNH S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 9 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Emergency Feedwater

System Emergency Feedwater Pumps 3 ASME III 3 I EF AE Piping and Valves 3 ASME III 3 I EF AE Fans 3 AMCA - I EF AE Ductwork 3 SMACNA - I EF AE Dampers 3 AMCA 6.8 Auxiliary Systems S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 10 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Spent Fuel Pool Cooling and

Cleanup System Spent Fuel Pool Cooling Pump 3 ASME III 3 I FB AE Spent Fuel Pool Skimmer Pump NNS MFRS. STDS. - - FB AE Spent Fuel Pool Heat

Exchanger 3 ASME III 3 I FB AE Spent Fuel Pool Demineralizer NNS ASME VIII - - PB AE Spent Fuel Pool Pre-Filter NNS MFRS. STDS. - - PB AE Spent Fuel Pool Post- Filter NNS MFRS. STDS. - - PB AE Spent Fuel Pool Skimmer NNS MFRS. STDS. - - FB AE Spent Fuel Pool Strainer NNS MFRS. STDS. - - FB AE Refueling Canal Skimmer Pump NNS MFRS. STDS. - - CS AE Piping and Valves Spent Fuel Pool Cooling

Loop 3 ASME III 3 I FB AE See Note 3. Spent Fuel Pool Skimmer

Loop NNS ANSI B31.1 - - FB/PB AE

9.1.3 Refueling

Canal Cleanup

Loop NNS ANSI B31.1 - - CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 11 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Station Service Water

System Service Water Pumps 3 ASME III 3 I CW AE Piping and Valves 3 ASME III 3 I SW/PB/YD AE

9.2.1 Strainers

3 ASME III 3 I PB AE Primary Component

Cooling Water System (PCCW) PCCW Pumps 3 ASME III 3 I PB AE PCCW Heat Exchanger 3 ASME III 3 I PB AE PCCW Head Tank 3 ASME III 3 I PB AE Piping and Valves Furnish and support cooling water supply to safeguards components 2/3 ASME III 2/3 I CS/CF/FS AE See Note 3. Containment Penetration 2 ASME III 2 I CE AE Other NNS ANSI B31.1 - - PB/CS WB/FB/YD AE Thermal Barrier Pumps 3 ASME III 3 I CS AE HX 2/3 ASME III 2/3 I CS AE 9.2.2 Head Tank 3 ASME III 3 I CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 12 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Demineralized Makeup

Water System Storage & Distribution Subsystem Demineralized Water Storage

Tanks NNS API-650, ASME VIII - YD AE Demineralized Water Transfer Pumps NNS MFRS. STDS. - - TB AE Demineralized Water Storage Tank Recirc. Pump NNS MFRS. STDS. - - CW NHY Demineralized Water Storage

Tank Heat Exchangers NNS ASME VIII - - TB/CW AE/NHY Piping and Valves NNS ANSI B31.1 - - PB/CD/TB AE Containment Penetration 2 ASME III 2 I CS AE 9.2.3 Water Treatment Subsystem NNS ASME VIII, MFRS. STDS. - - TB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 13 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Ultimate Heat Sink Atlantic

Ocean Intake & Discharge Tunnels NNS - - - - AE Piping 3 ASME III and 3 I YD AE NNS ANSI B31.1 - - YD/TB AE Expansion Joints 3 ASME III and - - YD AE NNS ANSI B31.1 3 1 PA/SW AE Cooling Tower Cooling Tower Fans 3 MFRS. STDS. - I CT AE Cooling Tower Pumps 3 ASME III 3 I CT AE 9.2.5 Piping and Valves 3 ASME III 3 I CT AE Condensate Storage Facilities Condensate Storage Tank 3 ASME III 3 I YD AE

9.2.6 Piping

and Valves NNS ANSI B31.1 - - YD/TB AE Reactor Makeup Water

System Reactor Makeup Water NNS API-650, - - YD AE Storage Tank ASME VIII Reactor Makeup Water Pump NNS ANSI B73.1 - - PB AE Piping and Valves Containment Penetration 2 ASME III 2 I CS AE 9.2.7 Other NNS ANSI B31.1 - - PB/WB/CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 14 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Compressed Air System Air Compressor NNS ASME VIII - - TB AE After-Cooler NNS ASME VIII - - TB AE Moisture Separator NNS ASME VIII - - TB AE Instrument Air Dryer Unit NNS ASME VIII - - TB AE Air Receiver NNS ASME VIII - - TB AE Filters NNS ASME VIII - - TB AE Piping and Valves Containment Penetration 2 ASME III 2 I CE AE Other NNS ANSI B31.1 - - A11 AE Vent Line from Safety-Related Devices to Class 1E Solenoid Vent Valve Tubing NNS ANSI B31.1 - I All AE Flexible Connectors NNS MFR STD - I All AE Safety-Related Backup Gas Supply Tubing and Valves 3 ANSI B31.1 - I All AE Flexible Connectors 3 ANSI B31.1 - I All AE Gas Cylinders 3 DOT 3AA - I All AE 9.3.1 Pressure Regulator 3 MFR STD - I All AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 15 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Process Sampling System Sample Heat Exchangers NNS ASME VIII - - PB AE Sample Pressure Vessels NNS ASME VIII - - PB AE Piping, Tubing and Valves Originating within Containment up to Isolation

Valve 2 ASME III 2 I CS/CE AE See Note 21. Post-Accident Sample Panel NNS No Code - - PB AE Post-Accident Sample Pumps NNS MFRS. STDS. - - PB AE 9.3.2 Other NNS ANSI B31.1 - - PB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 16 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Equipment and Floor Drainage System Reactor Coolant Drain Tank (RCDT) NNS ASME VIII - - CS AE RCDT Pump NNS No Code - - CS AE RCDT Heat Exchanger NNS ASME VIII - - CS AE Sump Pumps NNS No Code - - A11 AE Chemical Drain Tank NNS ASME VIII - - AB AE Chemical Drain Treatment

Tank NNS ASME VIII - - WB AE Chemical Drain Transfer Pump NNS No Code - - AB AE Chemical Drain Treatment Pump NNS No Code - - WB AE Piping and Valves Containment Penetration 2 ASME III 2 I CS AE 9.3.3 Other NNS ANSI B31.1 - - A11 AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 17 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Chemical and Volume

Control System Centrifugal Charging Pump 2 ASME III 2 I PB W See Notes 1a and 1c or 1d

and 2. Positive Displacement Pump 2 ASME III 2 I PB W See Notes 1a and 1c or 1d

and 2. Boric Acid Transfer Pump 3 ASME III 3 I PB W See Note 1b. Volume Control Tank 2 ASME III 2 I PB W Boric Acid Tank 3 ASME III 3 I PB AE Boric Acid Batching Tank NNS ASME VIII - - PB AE Chemical Mixing Tank NNS ASME VIII - - PB AE Resin Fill Tank NNS ASME VIII - - PB AE Regenerative Heat Exchanger 2 ASME III 2 I CS W Letdown Heat Exchanger PB W Tube Side 2 ASME III 2 I Shell Side 3 ASME III 3 I See Note 2. Excess Letdown Heat

Exchanger CS W Tube Side 2 ASME III 2 I Shell Side 3 ASME III 3 I Seal Water Heat Exchanger PB W Tube Side 2 ASME III 2 I Shell Side 3 ASME III 3 I See Note 2. Mixed Bed Demineralizer 3 ASME III 3 I PB AE 9.3.4 Cation Bed Demineralizer 3 ASME III 3 I PB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 18 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Boric Acid Blender 3 ASME III 3 I PB AE Batching Tank Agitator NNS MFRS. STDS. - - PB AE Letdown Degasifier NNS ASME VIII - - PB AE Letdown Degasifier NNS ASME VIII - - PB AE Preheater (Tube Side) Letdown Degasifier NNS ASME VIII - - PB AE Regenerative Heater Letdown Valves 2 ASME III 2 I CS AE Reactor Coolant Pump Seal 1 ASME III 1 I CS AE Bypass Orifice Boric Acid Transfer Pump 3 ASME III 3 I PB AE Bypass Orifice Chemical Mixing Tank

Orifice NNS ASME VIII - - PB AE Reactor Coolant Filter 2 ASME III 2 I PB AE Seal Water Injection Filter 2 ASME III 2 I PB AE Seal Water Return Filter 2 ASME III 2 I PB AE Boric Acid Filter 3 ASME III 3 I PB AE Demineralized Prefilter 3 ASME III 3 I PB AE Boron Meter NNS ANSI B31.1 - - PB W Classified on basis that flow restriction is provided in the piping.

9.3.4 (cont) Reactor Coolant Pump Standpipe NNS ASME VIII - - CS W

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 19 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Reactor Coolant Pump Standpipe Orifice NNS - - - CS W Chemical and Volume

Control System Piping 2 ASME III 2 I CS/PB W/AE See Note 3.

3 ASME III 3 I PB W/AE NNS ANSI B31.1 - - PB AE Valves 2 ASME III 2 I CS/PB W/AE See Note 3.

3 ASME III 3 I CS/PB W/AE NNS ANSI B31.1 - - CS/PB AE Boron Thermal

Regeneration Subsystem Chiller Pump NNS No Code - - PB W See Note 6.

Chiller Surge Tank NNS ASME VIII - - PB W See Note 6. Moderating Heat Exchanger PB W Tube Side 3 ASME III 3 -

See Note 6.

Shell Side 3 ASME III 3 -

See Note 6. Letdown Chiller Heat

Exchanger PB W Tube Side 3 ASME III 3 -

See Note 6.

Shell Side NNS ASME VIII - -

See Note 6. Letdown Reheat Heat

Exchanger PB W 9.3.4 (cont) Tube Side 2 ASME III 2 I See Note 6.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 20 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Shell Side 3 ASME III 3 -

See Note 6. Thermal Regeneration 3 ASME III 3 - PB W See Note 6. Demineralizer Chiller Unit W Evaporator NNS ASME VIII - -

See Note 6.

Condenser NNS ASME VIII - -

See Note 6.

9.3.4 (cont) Compressor NNS No Code - - See Note 6.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 21 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Boron Recovery System Recovery Evaporator Feed Pump NNS No Code - - WB AE Recovery Test Tanks NNS API-650, ASME VIII - - WB AE See Note 8. Recovery Test Tank Pump NNS No Code - - WB AE Boron Waste Storage Tanks NNS API-620, ASME VIII - - WB AE See Note 8. Recovery Evaporator Package Recovery Evaporator NNS ASME VIII - - WB AE See Note 8. Bottoms Pump NNS No Code - - WB AE See Note 8. Bottoms Cooler NNS ASME VIII - - WB AE See Note 8.

Reboiler NNS ASME VIII, TEMA C - - WB AE See Note 8. Reboiler Pump NNS No Code - - WB AE See Note 8. Distillate Pump NNS No Code - - WB AE See Note 8. Distillate Accumulator NNS ASME VIII - - WB AE See Note 8. Distillate Condenser NNS ASME VIII - - WB AE See Note 8.

Distillate Cooler NNS ASME VIII, TEMA C - - WB AE See Note 8. Cesium Removal Ion

Exchanger NNS ASME VIII - - WB AE See Note 8. Recovery Demineralizer NNS ASME VIII - - WB AE See Note 8. Primary Drain Tank (PDT) NNS ASME VIII - - WB AE PDT Transfer Pump NNS MFRS. STDS. - - WB AE 9.3.5 PDT Degasifier NNS ASME VIII - - WB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 22 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) PDT Degasifier Recirc. Pump NNS MFRS. STDS. - - WB AE PDT Degasifier Preheater NNS ASME VIII, TEMA R - - WB AE PDT Degasifier Regenerative

Heat Exchanger NNS ASME VIII, TEMA R - - WB AE PDT Degasifier Trim Cooler NNS ASME VIII, TEMA R - - WB AE Filters NNS ASME VIII - - WB AE 9.3.5 (Cont) Piping and Valves NNS ANSI B31.1 - - WB AE Equipment Vent System Safety Valve Surge Tank NNS ASME VIII - - PB AE Evacuation Pump NNS MFRS. STDS. - - CS PSNH Separator/Silencer NNS ASME VIII - - CS PSNH Piping and Valves Containment Penetration 2 ASME III 2 I CS AE 9.3.6 Other NNS ANSI B31.1 - - WB/PB/CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 23 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Control Room Complex

Ventilation System Control Room Air Conditioning Subsystem Air Conditioning/Fan 3 MFRS. STDS. - I CD AE Water Chiller 3 MFRS. STDS. - I CD AE Dampers 3 AMCA - I CD AE Ductwork 3 SMACNA - I CD AE See Note 13. Computer Room Air Conditioning Subsystem Air Conditioning Unit - MFRS. STDS. - - CD AE Condensing Unit - MFRS. STDS. - - CD AE Cooling Coil 3 ARI 410 - I CD Chilled Water Pumps 3 MFRS. STDS. - I CD AE Expansion Tank 3 ASME VIII - I CD AE Piping and Valves 3 ANSI B31.1 - I CD AE Dampers - AMCA - - CD AE Ductwork - SMACNA - - CD AE See Note 20. Control Room Complex Normal Makeup Air Subsystem Fans 3 AMCA - I CD AE Dampers 3 AMCA - I CD AE 9.4.1

Intake Piping 3 ANSI B31.1 NNS__

1A I CD/YD AE See Note 19.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 24 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) 9.4.1 (Cont.) Control Room Emergency Filtration Subsystem Fuel Storage Building

Ventilation System Ventilation Fans - AMCA - - FB AE Dampers 3 AMCA - I FB AE 9.4.2 Ductwork - SMACNA - I FB AE Primary Auxiliary Building

Ventilation System Ventilation Fans AMCA - - PB AE PCCW and BI Pump Area Wall AMCA - I PB AE Fans and Dampers Normal Cleanup Exhaust Fans AMCA - - PB AE Normal Cleanup Filters ANSI N509/

N510 - - PB AE Ductwork and Dampers Isolation Dampers and Ductwork Passing through CE 2 ANSI N509/N510 - I PB/CE AE

9.4.3 Other

NNS SMACNA - I PB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 25 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Waste Processing Building

Ventilation System Fans - AMCA - - WB AE Ductwork - SMACNA - - WB AE Dampers - AMCA - - WB AE 9.4.4 Filters - ANSI N509/N510 - - WB AE Containment Ventilation

System Recirculation Fan/Coolers - MFRS. STDS. - I CS AE See Note 9.

9.4.5 Ductwork

- SMACNA - I CS AE See Note 10.

Containment Enclosure

Ventilation System Fans 3 AMCA - I CE AE Cooling Coils 3 ASME III - I CE AE 9.4.6 Ductwork 3 ANSI N509/ N510 - I CE AE Electrical Penetration Area

Air Conditioning System Air Conditioning Unit - MFRS. STDS. - - EF AE Condensing Unit - MFRS. STDS. - - EF AE Ductwork - SMACNA - - EF AE 9.4.7 Dampers - AMCA - - EF AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 26 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Diesel Generator Building

Ventilation System Fans 3 AMCA - I CD AE Ductwork 3 (Mod) SMACNA - I CD AE

9.4.8 Dampers

3 AMCA - I CD AE Cable Spreading Room

Ventilation System Fans - AMCA - - CD AE Ductwork - SMACNA - I CD AE 9.4.9 Dampers - AMCA - I CD AE 4 kV Switchgear Area and Battery Rooms Ventilation

System Fans 3 AMCA - I CD AE Dampers 3 AMCA - I CD AE 9.4.10 Ductwork 3 (Mod) SMACNA - I CD AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 27 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Emergency Feedwater

Pumphouse Ventilation

System Fans 3 AMCA - I EF AE Dampers 3 AMCA - I EF AE 9.4.11 Ductwork 3 (Mod) SMACNA - I EF AE Service Water Pumphouse

Ventilation System Fans 3 AMCA - I CW AE Dampers 3 AMCA - I CW AE 9.4.13 Ductwork 3 (Mod) SMACNA - I CW AE Service Water Cooling

Tower Pump Room and Switchgear Room

Ventilation Systems Pump Room Power Roof Ventilator 3 AMCA - - CT AE Switchgear Room Supply Fan 3 AMCA - I CT AE Ductwork 3 (Mod) SMACNA - I CT AE 9.4.14 Dampers 3 AMCA - I CT AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 28 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Fire Protection System Water Supply Tank - NFPA STDS. - - YD AE Fire Pump - NFPA STDS. - - YD AE Jockey Pump - NFPA STDS. - - YD AE

9.5.1 Piping

and Valves - NFPA STDS. - - YD AE Diesel Generator Fuel Oil Storage and Transfer

System Fuel Oil Day Tank 3 ASME III 3 I CD AE Fuel Oil Storage Tank 3 ASME III 3 I CD AE Fuel Oil Transfer Pump 3 ASME III 3 I CD AE Associated Piping and Valves for Above Tanks and Pump 3 ASME III 3 I CD AE 9.5.4 All Remaining On-Engine Equipment and Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 29 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Diesel Generator Cooling

Water System Auxiliary Coolant Pump 3 MFRS. STDS. - I CD AE DG Component Cooling Water Heat Exchanger 3 ASME III 3 I PB AE Expansion Tank 3 ASME III 3 I CD AE Interconnecting Piping and Valves for Above Components, and On-Skid, Off-Engine Piping 3 ASME III 3 I CD/PB/YD AE 9.5.5 All Remaining Essential On-Engine Equipment and Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.

Diesel Generator Starting

System Air Receivers 3 ASME III 3 I CD AE Air Compressor 3 MFRS. STDS. - I CD AE Piping and Valves from Air

Receiver to DG Skid 3 ASME III 3 I CD AE All Remaining Essential Equipment Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.

9.5.6 Backup

Control Air Compressor 3 MFRS. STDS. - I CD S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 30 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Diesel Generator

Lubrication System Auxiliary Lube Oil Pump 3 MFRS. STDS. - I CD AE Associated Piping and Valves for Above Pump 3 ASME III 3 I CD AE 9.5.7 All Remaining Essential On-Engine Equipment and Piping 3 MFRS. STDS. - I CD AE See system P&ID for safety class boundaries.

Diesel Generator

Combustion Air Intake and

Exhaust System Piping (Off-Engine) 3 ANSI B31.1 - I CD AE Air Intake Filter 3 MFRS. STDS. - I CD AE Exhaust Silencer 3 MFRS. STDS. - I CD AE Exhaust Expansion Joints 3 MFRS. STDS. - I CD AE Piping (On-Engine) 3 MFRS. STDS. - I CD AE

9.5.8 Steam

& Power Conversion System S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 31 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Main Steam System Main Steam Piping and Valves from Steam Generator up through Containment Isolation Valves 2 ASME III 2 I MF/CS AE To Emergency Feedwater Pump Turbine Interface 3 ASME III 3 I MF/EF AE EFW Pump Turbine Assembly 3 MFRS. STDS. - I MF/EF AE 10.3 Other NNS ANSI B31.1 - - TB AE Main Condenser Evacuation System Vacuum Pump NNS MFRS. STDS. - - TB AE 10.4.2 Piping and Valves NNS ANSI B31.1 - - TB AE Turbine Gland Sealing

System 10.4.3 Gland Steam Condenser and Exhauster Fans NNS MFRS. STDS. - - TB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 32 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Condensate and Feedwater

System Condensate Pump NNS MFRS. STDS. - - TB AE Heater Drain Pump NNS MFRS. STDS. - - TB AE Steam Generator Feed Pump NNS MFRS. STDS. - - TB AE Steam Generator Startup Feed Pump NNS MFRS. STDS. - - TB AE Feedwater Heater NNS ASME VIII - - TB AE Piping and Valves From First Isolation Valve Outside Containment to Steam Generator 2 ASME III 2 I MF/CS AE 10.4.7 Other NNS ANSI B31.1 - - TB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 33 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Steam Generator Blowdown

System Flash Tank NNS ASME VIII - - PB AE Flash Tank Bottoms Cooler NNS ASME VIII, TEMA C - - PB AE Flash Steam Condenser/

Cooler NNS ASME VIII, TEMA C - - PB AE Flash Tank Distillate Pump NNS MFRS. STDS. - - PB AE Blowdown Evaporator PB AE See Note 8. Vapor Body NNS ASME VIII - -

Heating Element NNS ASME VIII, TEMA C - - Distillate Condenser NNS ASME VIII, TEMA C - - Distillate Accumulator NNS ASME VIII, TEMA C - - Distillate Pump NNS MFRS. STDS. - -

Distillate Cooler NNS ASME VIII, TEMA C Bottoms Pump NNS MFRS. STDS. - -

Bottoms Cooler NNS ASME VIII - -

10.4.8

Piping and Valves Inside Containment, up through Isolation Valves 2 ASME III 2 I CS AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 34 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) 10.4.8 (Cont.) Other Piping and Valves NNS ANSI B31.1 - - PB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 35 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Radioactive Waste Management Liquid Waste System Waste Evaporator Package Reboiler NNS ASME VIII, - - WB AE See Note 8.

TEMA C Reboiler Pump NNS MFRS. STDS. - - WB AE See Note 8. Bottoms Pump NNS MFRS. STDS. - - WB AE See Note 8. Bottoms Cooler NNS ASME VIII - - WB AE See Note 8. Waste Evaporator NNS ASME VIII - - WB AE See Note 8. Distillate Pump NNS MFRS. STDS. - - WB AE See Note 8. Distillate Accumulator NNS ASME VIII - - WB AE See Note 8.

Distillate Cooler NNS ASME VIII, TEMA C - - WB AE See Note 8. Distillate Condenser NNS ASME VIII, TEMA C - - WB AE See Note 8.

Floor Drain Tank NNS ASME VIII - - WB AE See Note 8. Waste Test Tank NNS ASME VIII - - WB AE See Note 8.

Duplex Strainer NNS MFRS. STDS. - - WB AE Waste Test Tank Pump NNS MFRS. STDS. - - WB AE Floor Drain Tank Pump NNS MFRS. STDS. - - WB AE Waste Demineralizer NNS ASME VIII - - WB AE See Note 8. Waste Demineralizer Filter NNS ASME VIII - - WB AE Floor Drain Tank Filter NNS ASME VIII - - WB AE 11.2 Piping and Valves NNS ANSI B31.1 - - WB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 36 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Gaseous Waste System Carbon Delay Beds NNS ASME III - - WB AE Iodine Guard Beds NNS ASME III - - WB AE Gas Chiller NNS ASME III - - WB AE Waste Gas Dryer NNS ASME III - - WB AE Hydrogen Surge Tank NNS ASME VIII - - WB AE Purge Gas Condenser NNS ASME VIII - - WB AE Gas Chiller Compressor Unit NNS MFRS. STDS. - - WB AE Hydrogen Gas Compressor

Package NNS MFRS. STDS. - - WB AE Regenerative Compressor

Unit Package NNS MFRS. STDS. - - WB AE Piping and Valves in process portion of piping up to and including the carbon delay

beds NNS ASME III - - WB AE See Note 14. Carbon delay bed support element up to and including the attachment weld NNS AISC - I WB AE See Note 17.

11.3 Other Piping and Valves NNS ANSI B31.1 - - WB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 37 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Solid Waste System Spent Resin Sluice Tank NNS ASME VIII - - WB AE Spent Resin Sluice Pump NNS MFRS. STDS. - - WB AE Spent Resin Sluice Filter NNS ASME VIII - - WB AE Spent Resin Transfer Pump NNS MFRS. STDS. - - WB AE Waste Concentrates Tank NNS ASME VIII - - WB AE Waste Concentrates Transfer Pump NNS MFRS. STDS. - - WB AE Resin Hopper NNS ASME VIII - - WB AE Spent Resin Dewatering Pump NNS MFRS. STDS. - - WB AE Resin Centrifuge Metering Pump NNS MFRS. STDS. - - WB AE Resin Centrifuge NNS MFRS. STDS. - - WB AE Crystallizer Skid Vessels NNS ASME VIII - - WB AE Heat Exchangers NNS TEMA R - - WB AE Pump NNS MFRS. STDS. - - WB AE Extruder/Evaporator NNS DIN - - WB AE Asphalt Storage Tank NNS API 650 - - WB AE Auxiliary Boiler Vessel NNS ASME VIII - - WB AE Asphalt Recirculation Pump NNS MFRS. STDS. - - WB AE Asphalt Metering Pump NNS MFRS. STDS. - - WB AE 11.4 Extruder Vent Hood NNS - - WB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 38 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Auxiliary Boiler Chemical

Feed Skid NNS MFRS. STDS. - - WB AE S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 39 of 42 UFSAR Section Systems and Components ANS Safety Class Principal Design/Const.

Codes/Stds.

Code Class Seismic Category Building (11) Supplier Notes (14) Auxiliary Boiler Blowdown

Tank NNS MFRS. STDS. - - WB AE Steam Condensate Collection

Skid NNS MFRS. STDS. - - WB AE Auxiliary Boiler Feed Pumps NNS MFRS. STDS. - - WB AE Auxiliary Boiler NNS MFRS. STDS. - - WB AE Condensate Return Tank Trash Compactor NNS MFRS. STDS. - - WB AE Trash Compactor Filter NNS MFRS. STDS. - - WB AE Trash Compactor NNS MFRS. STDS. - - WB AE 11.4 (Cont.) Hydraulic Unit Piping and

Valves NNS ANSI B31.1 - - WB AE

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 40 of 42 NOTES 1. Services required to support a sa fety or other n ecessary function: a. Emergency power - automatic loading. b. Emergency power - manual loading. c. Component cooling water. d. Service water.

2. Portions of equipment containing component cooling water are SC-3, Code Class 3. 3. Safety classes for piping and valves are as defined by the sy stem piping and instrumentation diagrams. Code classes are those required by the safety class. 4. Represents code class upgrading. As perm itted by paragraph NA-2134 of the ASME Code,Section III, this component is upgraded from the minimum required Code Class 2 to Code Class 1. 5. Parts are mechanically of safety class and must meet the structural integrity requirements of the specification and quality assurance requirements of 10 CFR 50, Appendix B. 6. This component is Safety Class 3 unde r the definition 2.2.3(1), (3), or (4) of ANSI N18.2-1975 and qualifies for no special seismic design by meeting the four following conditions. Portions of systems in which this component is located that perform the same safety function likewise qualify for no special seismic design. Conditions to be met for exemption are the following: a. Failure would not directly cause a Cond ition III or IV event (as defined in "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," ANSI N18.2-1973). b. There is no safety function to mitigate nor could failure prevent m itigation of th e consequences of a Condition III or IV ev ent. c. Failure during or following any Condition II event would result in consequences no more severe than allowed for a Condition III event. d. Routine post-seismic procedures would disclose loss of the safety function S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 41 of 42 7. ASME: American Society of Mechanical Engineers. III stands for Section III of the ASME Boiler and Pressure Vessel Code (B&

PVC). VIII stands for Section VIII of the ASME B&PVC. Pressure vessels which ar e part of the reactor coolant pressure boundary meet the requirements of the 1971 Version, with application of all Addenda through to and including the Summer 1972 Addenda. Pumps, valves and piping which are part of the reactor coolant pressure boundary meet the requirements of the 1971 Version, with application of all Addenda through to and including the Winter 1972 Addenda. Later code versions may be used optionally. 8. Augmented to provide additional quality assurance. Provisions include completely welded systems except where maintenance or testing requires flanged connections, material certifications consistent with ASME Section III, ND-2121, and mandatory hydrostatic testing of all systems. 9. Fan/cooling units are designed and seismically analyzed to a ssure that they will not overturn or fail structurally during an SSE. 10. Ductwork supports are designed and seismically analyzed to withstand an SSE where support failure could damage safeguards equipment. 11. Building code: AB = Administration and Service Building CE = Containment Enclosure Building CD = Control and Diesel Generator Building CS = Containment Structure CT = Service Water Cooling Tower CW = Service and Circulating Water Pumphouse EF = Auxiliary Feedwater House & Electrical Penetration Area FB = Fuel Storage Building PB = Primary Auxiliary Building MF = Main Steam and Feedwater Pipe Chase TB = Turbine Building WB = Waste Processing Building YD = Yard Arrangement drawings for the buildings in which the systems are located are presented in Section 1.2.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , AND SYSTEMS TABLE 3.2-2 Revision:

Sheet: 13 42 of 42 12. Ductwork from the downstream side of the air cleaning units to the fan intakes and discharge of the fans to the building boundaries is Safety Class 3, seismic Category I. 13. Ductwork located within the mechanical equipment room to the boundary of the control room is Safety Class 3, seismic Catego ry I 14. Motors, valve operators and valve actuators which must operate (run, open or close) in order for the system to perform its safety function are classified as within the scope of the OQAP. Motors or operators which are associated with mechanical components which serve only as part of a pressure boundary are not within the scope of the OQAP. 15. Deleted 16. (Deleted in Amendment 60) 17. The tank support elements should satisfy the requireme nts of Position 5 of Regulatory Guide 1.143, Rev. 1. 18. System specifications provided under Updated FSAR Subsection 6.5.1. 19. Remote air intake piping m eets ANS/HVAC Safety Class 3 crite ria and NNS-1A piping criteria. 20. Control room air conditioning system provides cooling air to the computer room in the event computer room air conditioning system fails. 21. Components will have same safety class/

codes as process piping if connected to Safety Class 3 or lo wer classifications. 22. The CAP penetrations (HVAC-1, HVAC-2) a nd associated blind flanges are qualified to , and classified as, ASME III, subsectio n NE, Class MC.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , COMPONENTS , EQUIPMENT , AND SYSTEMS TABLE 3.2-3 Revision:

S HEET: 15 1 of 1 TABLE 3.2-3 QUALITY STANDARDS (1) Components(2) Safety Class 1 Safety Class 2 Safety Class 3 Non Nuclear Safety (NNS) Pressure Vessels ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, Class 1 ASME Boiler and Pressure

Vessel Code,Section III, Nuclear Power Plant Components, Class 2 ASME Boiler and Pressure

Vessel Code,Section III, Nuclear Power Plant Components, Class 3 ASME Boiler and Pressure

Vessel Code,Section VIII, Division 1 Piping As above As above As above ANSI B31.1.0 Power Piping Pumps As above As above As above ASME Boiler and Pressure Vessel Code, Sec. VIII, Div.

1 and Mfrs. Standards Valves As above As above As above ANSI B31.1.0 Atmospheric Storage

Tanks - As above As above API-650 0-15 psig Storage

Tanks - As above As above API-620 Piping used as HVAC

Duct - - ANSI B31.1.0 ANSI B31.1.0 (1) Pressure vessels which are part of the reactor coolant pressure boundary meet the requirements of the 1971 Version of the ASME Boiler and Pressure Vessel Code, w ith application of all addenda through to and including the Summer 1972 Addenda. Pumps, valves and piping which are part s of the reactor coolant pr essure boundary meet the requirements of the 1971 Version, with application of all addenda through to and including the Winter 1972 Addenda. Later code versions may be used optionally. (2) Components other than those described in this table are controlled by quality standards discussed in the component descriptions contained in pertinent sections of the UFSAR.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT AND SYSTEMS TABLE 3.2-4 Revision:

Sheet: 8 1 of 1 TABLE 3.2-4 S EISMIC A ND S AFETY C LASSIFICATIONS H VAC S YSTEMS A ND C OMPONENTS Ductwork Supports Function Classification Symbol Criteria Guideline Criteria Guideline Safety- Related Class 2 Seis. Cat. I 2 (Note 1) Nonrigid Allowed Rigid Nonrigid Allowed Rigid Class 3 Seis. Cat. I 3* (Note 2) Nonrigid Allowed Rigid Nonrigid Allowed Rigid Nonsafety-Related Seismic Category I 1 Nonrigid Allowed Rigid SMACNA (Note 3) Nonrigid Allowed Rigid Nonseismic Nonsafety - - SMACNA (Note 3) - SMACNA Notes 1. These are 10 gauge welded ductwork leak tested to ANSI N509/N510 requirement. 2. These are either 10 gauge welded ductwork leak tested to ANSI N509/N510 requirements or in some specific instances upgraded SMACNA high pressure ductwork leak tested to SMACNA requirements. 3. These are standard SMACNA high pressure ductwork leak tested to SMACNA requirements.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , A ND SYSTEMS TABLE 3.3-1 Revision:

Sheet: 8 1 of 1 TABLE 3.3-1 W IND P RESSURE ON CONTAINMENT E NCLOSURE B UILDING Height Above Wind Velocity q (1) p = 0.5q (2) Grade (feet)

(mph) (psf) (psf) 0 - 50 110 37 19 50-150 135 56 28 Above 150 165 84 42 (Reference - ASCE Paper 3269) (1) q - .00256 (1.1V) 2 (2) The full value of p is distributed horizontally by means of the pressure coefficients (CP) on Figure 3.3-1 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS , E QUIPMENT , A ND SYSTEMS TABLE 3.3-2 Revision:

Sheet: 8 1 of 1 TABLE 3.3-2 W IND VELOCITY P RESSURES Height of Grade q F q p q M PSF PSF PSF 30' or less 40 46 31 Over 30' and up to 50' 46 51 36 Over 50' and up to 100' 53 59 44 Over 100' and up to 150' 58 65 49 Over 150' and up to 200' 62 69 53 Over 200' and up to 250' 65 72 57 (Reference - ANSI-A58.1-1972)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.3-3 Revision:

Sheet: 8 1 of 1 TABLE 3.3-3 T ORNADO W IND P RESSURE C OMPONENTS ASCE ANSI W W1 = External Tornado Wind Pressure For Rectangular Structures Wind Wall 0.8 q avg. Leeward Wall

-0.5 q avg. Side Walls

-0.7 q avg Roof -0.7 q avg. For Cylindrical Structures:

Cpe (2) x q avg.(1) Cpe x q avg. W w2 = Internal Tornado Wind Pressure

+/-0.2 (3) q avg. +/-0.3 q avg. (1) q avg. = C s qmax. = C s (332) PSF. See Fig. 3.3-2 for C s (2) C pe = External Pressure Coefficient. See Fig. 3.3-1 for C pe (3) Coefficient +/-0.2 is for uniformly distributed openings. All other cases are evaluated from Table 4.0 of Ref. 1

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.3-4 Revision:

Sheet: 8 1 of 1 TABLE 3.3-4 N ON-C ATEGORY I S TRUCTURES D ESIGNED AGAINST COLLAPSE O NTO A DJACENT C ATEGORY I S TRUCTURES D UE T O T ORNADO W IND L OADING Non-Category I Structures Affected Category I Structure or Component Turbine Building (1) Condensate Storage Tank (E.W.), Containment and Others (N.S.) Nonessential Switch gear Building (2) Control and Diesel Generator Building Tank Farm Area (Steel Framing Portion which includes Steel Framing, Concrete Roofing and Metal Siding over Refueling Water Storage Tank)

(3) Primary Auxiliary Building, Waste Processing Building and Tank Farm Area Tunnels Steam Generator Blowdown Recovery Building (4) Primary Auxiliary Building and Waste Process Building Circulating Water Pumphouse Steel Framing Portion (5) Service Water Pumphouse Waste Processing Building Steel Framing Portion and

reinforced concrete portion, except the area between Columns 1 & 2 and A to D between elevations 53'-0" and 86'-0" Primary Auxiliary Building, Tank Farm Area and Piping Tunnels NOTES: (1) The entire Turbine Building is designed against failure in the north-south direction. The south end is designed against failure in the east-west direction; an east-west failure in the north end will not affect any seismic Category I structures.

(2) The Nonessential Switchgear Building is designed mechanistically to fall away from the Control and Diesel Generator Building under the action of a collapsing Administration and Service Building. Thus, no significant load is applied to the Control and Diesel Generator Building by either the falling Administration and Service Building or the falling Nonessential Switchgear Building. (3) The steel framing portion has been evaluated for tornado wind loads against collapse on surrounding safety-related systems, components and structures and it has been found that there are no adverse effects. The tornado effects of the steel framing portion upon the systems and components located within the tank farm structure are

not a design consideration because the loss of function of these systems and components will not affect the capability of a safe reactor shutdown.

(4) The collapse of the steel framing upon surrounding Category I structures and components as stated above was evaluated to prove that these category structures and components are not damaged to the extent of not being able to perform their function. The tornado effects of the steel framing portion upon the systems and components located within the Steam Generator Blowdown Recovery Building are not a design consideration because the loss of function of these systems and components will not affect the capability of a safe reactor shutdown.

(5) The collapse of the Circulating Water Pumphouse on the Service Water Pumphouse was evaluated to prove that the collapse will not impair the Service Water Pumphouse or system.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.5-1 Revision:

Sheet: 8 1 of 1 TABLE 3.5-1 S TRUCTURES , SHIELDS A ND B ARRIERS D ESIGNED T O R ESIST I NTERNAL A ND EXTERNAL M ISSILES Internal Missiles External Missiles Containment shell 1 Containment structure primary shield wall 2 Secondary shield wall 2 Pressurizer shielding 2 Reactor missile shield 2 Operating floor 2 Primary Auxiliary Building 3 1 Refueling water storage tank See Section

6.2 Condensate

storage tank 1 Control and Diesel Generator Building 3 1 Shield between diesel generators 3 Fuel Storage Building 3 1 Electrical penetration tunnel 1 Service Water Pumphouse, intake and discharge transition structures, flume and backwash ducts 3 1 Emergency Feedwater Pump Building 3 1 Diesel fuel tanks Piping tunnels 3 1 Main steam and feedwater lines outside containment to 2nd isolation valve 1 Category I electrical manholes 1 1 - Tornado generated missiles 2 - Missiles postulated within the containment 3 - Temperature sensing element missile

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-2 Revision:

Sheet: 8 1 of 1 TABLE 3.5-2 S UMMARY O F C ONTROL R OD D RIVE M ECHANISM M ISSILE A NALYSIS Weight Thrust Area Impact Area Impact Velocity Kinetic Energy Postulated Missiles (lb) (in.2) (in.2) (ft/sec) (ft-lb) Mechanism Housing Plug 6.5 a 5.94 7.07 255 6563 Drive Shaft 170 b 2.40 2.41 97.6 25147 a. Reflects Removal of Eyebolt

b. Heavy Drive Rod.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-3 Revision:

Sheet: 8 1 of 1 TABLE 3.5-3 V ALVE - M ISSILE C HARACTERISTICS Missile Description Weight (lb.) Flow Discharge Area (in 2) Thrust Area (in 2) Impact Area (in 2) Weight to Impact Area Ratio (psi)

Velocity (fps) Safety Valve Bonnet 350 2.86 80 24 14.6 110 3 Inch Motor-Operated Isolation Valve Bonnet (plus motor operator and stem) 400 5.5 113 28 14.1 135 3 Inch Air-Operated Valve Bonnet (plus stem) 75 1.8 20 20 3.75 115 4 Inch Air-Operated Spray

Valve Bonnet 200 9.3 50 50 4 190 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.5-4 Revision:

Sheet: 8 1 of 1 TABLE 3.5-4 PIPING TEMPERATURE E LEMENT A SSEMBLY - M ISSILE C HARACTERISTICS

1. For a tear around the weld between the boss and the pipe:

Characteristics "Without well" "With well" Flow Discharge Area 0.11 in 2 0.60 in 2 Thrust Area 7.1 in 2 9.6 in 2 Missile Weight 11.0 lb. 15.2 lb. Area of Impact 3.14 in 2 3.14 in 2 Missile Weight Impact Area 3.5 psi 4.84 psi Velocity 20 ft/sec 120 ft/sec 2. For a tear at the junction between the temperature element assembly and the boss for the "without well" element and at the junction betw een the boss and the well for the "with well" element. Characteristics "Without well" "With well" Flow Discharge Area 0.11 in 2 0.60 in 2 Thrust Area 3.14 in 2 3.14 in 2 Missile Weight 4.0 lb. 6.1 lb. Area of Impact 3.14 in 2 3.14 in 2 Missile Weight Impact Area 1.27 psi 1.94 psi Velocity 75 ft/sec 120 ft/sec

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.5-5 Revision:

Sheet: 8 1 of 1 TABLE 3.5-5 C HARACTERISTICS O F O THER M ISSILES P OSTULATED W ITHIN R EACTOR C ONTAINMENT Reactor Coolant Pump Temperature Element Instrument Well of Pressurizer Pressurizer Heaters Weight 0.25 lb. 5.5 lb. 15 lb. Discharge area 0.5 in 2 0.442 in 2 0.6 in 2 Thrust area 0.5 in 2 1.35 in 2 2.4 in 2 Impact area 0.5 in 2 1.35 in 2 2.4 in 2 Missile Weight Impact Area 0.5 psi 4.1 psi 6.25 psi Velocity 260 ft/sec 100 ft/sec 55 ft/sec

S EABROOK D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT Revision:

8 S TATION A ND SYSTEMS Sheet: 1 of 1 UFSAR TABLE 3.5-6 TABLE 3.5-6 H YPOTHETICAL M ISSILE DIMENSIONS S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-7 Revision:

Sheet: 8 1 of 1 TABLE 3.5-7 H YPOTHETICAL M ISSILE D ATA Missile No. of Type of Wheel Group 1-3 Wheel Group 4-6 Wheel Group 7 Type Missiles Overspeed (%) Weight (lb.) Energy (106ft lb.)

Velocity (ft/sec) Weight (lb.) Energy (106ft lb.)

Velocity (ft/sec) Weight (lb.) Energy (106ft lb.)

Velocity (ft/sec) a 2 180 120 2000 0-8 0-510 4000 0-17 0-520 8200 26-53 10-22 450-650 280-420 b 1 180 120 1000 0-8 0-720 2000 0-16 0-720 4100 0-38 0-18 0-770 0-530 c 3 180 120 300 0-5 0-1040 600 0-8 0-930 1400 0-16 0-8 0-860 0-610 d 10 180 120 100 0-2 0-1130 150 0-2 0-930 200 0-3 0-2 0-980 0-800 Wheel Group 1-3 Wheel Group 4-6 Wheel Group 7 Missile Exit Angle a -5 to + 5 degrees -5 to + 5 degrees 0 to 25 degrees a - From Reference 4

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-8 Revision:

Sheet: 8 1 of 1 TABLE 3.5-8 D ATA O N S AFETY-R ELATED S TRUCTURES UNIT BUILDING NO. BUILDING NAME WALL THICKNESS (ft) ROOF THICKNESS (ft) EFFECTIVE ELEVATION MINIMUM TARGET (ft) MAXIMUM SIZE CONCRETE STRENGTH (psi) 1 1 Containment 4.5 - 20 119 74.25' radius 3000 1 2 Containment 3.5 3.5 119 183 74.25' radius 3000 1 3 Emergency Feedwater Pump 2.0 2.0 47 47 35'x82' 3000 1 4 Fuel Storage Building 2.0 2.0 20 84 98'x107.5' 3000 1 5 Primary, Auxiliary Bldg.

2.0 2.0 20 100 63.5'x145' 3000 1 6 Control Building 2.0 2.0 20 98 90'x138' 3000 1 7 Diesel Generator Bldg.

2.0 2.0 20 78.5 90'x93' 3000 1 8 Condensate Storage Tank 2.0 0.0 20 47.5 23.5' radius 3000 1 9 Equipment Vaults 2.5 2.5 2.5 53 41'x41.5' 3000 1 10 Service Water Pumphouse 2.0 2.0 20 50.25 60'x159' 3000 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-9 Revision:

Sheet: 10 1 of 1 TABLE 3.5-9 PROBABILITY O F D AMAGE B Y A L OW T RAJECTORY M ISSILE STRIKE O N B UILDINGS I N D IRECT Z ONE PROBABILITY OF DAMAGE PER TURBINE FAILURE (P2P3) (x10

-3 events/turbine failure) 120% Overspeed P1 = 1.1x10

-6/year 180% Overspeed P1 = 1. 1x10

-6/year Building Turbine 1 Turbine 2 Total Turbine 1 Turbine 2 Total Total Prob. of Damage (P1P2P3) (x10-7 events/yr) Containment - - - - - - - Control Building - - - - - - - Diesel Generator Bldg. 0.86 - 0.86 1.00 - 1.00 0.021 Condensate Storage Tank - - - - - - - Emerg. Feedwater Pump Bldg. - - - - - - - Total 0.86 - 0.86 1.00 - 1.00 0.021 Service Water Pumphouse - - - - - - -

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-10 Revision:

Sheet: 10 1 of 1 TABLE 3.5-10 PROBABILITY O F A HIGH T RAJECTORY M ISSILE STRIKE - R OOF H ITS O NLY PROBABILITY OF DAMAGE PE R TURBINE FAILURE (P2) (x10-3 events/turbine failure) 120% Overspeed P1 = 3.7x10

-5/year 180% Overspeed P1 = 3.7x10

-5/year Building Turbine 1 Turbine 2 Total Turbine 1 Turbine 2 Total Total Prob. of Strike (P1P2) (x10-7 events/yr) Containment .045 -

.045 .576 - .576 .230 Control Building

.032 - .032 .414 - .414 .165 Diesel Generator Bldg. .026 - .026 .273 - .273 .111 Emerg. Feedwater Pump Bldg. .007 - .007 .094 - .094 .037 Equipment Vault

.004 - .004 .055 - .055 .022 Primary Auxiliary Bldg. .023 - .023 .258 - .258 .104 Fuel Storage Building .026 - .026 .217 - .217 .090 Service Water Pumphouse .023 - .023 .194 - .194 .080 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.5-11 Revision:

Sheet: 8 1 of 1 TABLE 3.5-11 T ORNADO-G ENERATED M ISSILES A ND VELOCITIES MISSILE, DESCRIPTION FRACTION OF TOTAL TORNADO VELOCITY 1 HORIZONTAL VELOCITY IN REGION I (fps) A. Wood plank, 4 in. x 12 in. x 12 ft weight 200 lb. 0.8 422 B. Steel pipe, 3 in. diameter, schedule 40 10 ft long, weight 78 lb.

0.4 211 C. Steel rod, 1 in. diameter x 3 ft long weight 8 lb.

0.6 317 D. Steel pipe, 6 in. diameter, schedule 40 15 ft long, weight 285 lb.

0.4 211 E. Steel pipe, 12 in. diameter, schedule 40 15 ft long, weight 743 lb.

0.4 211 F. Utility pole, 131/2 in. diameter, 35 ft long, weight 1490 lb.

0.4 211 G. Automobile, frontal area 20 ft 2 weight 4000 lb.

0.2 106 1 The maximum wind speed in Region is 360 MPH (528 fps) per Regulatory Guide 1.76, Design Basis Tornado for Nuclear Power Plants.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.5-12 Revision:

Sheet: 8 1 of 1 TABLE 3.5-12 STRUCTURES AND BARRIER S DESIGNED TO RESIST TORNADO-GENERATED MISSILES* Containment Structure Containment Enclosure Ventilation Area Containment Equipment Hatch Missile Shield Control and Diesel Generator Building Control Room Makeup Air Intake Structure Dikes and Foundation for Refueling Water Storage Tank and Reactor Makeup Water Storage Tank, walls E.7-D.1 @ Col. line 0.5 and E-5 @ Col. line 4.5 East Penetration Area Emergency Feedwater Pump Building Enclosure for Condensate Water Storage Tank, vertical walls Fuel Storage Building, except rolling steel door in external wall on Col. line A Valve Pit Areas of the Intake and

Discharge Transition Structure Main Steam and Feedwater Pipe Chase (E&W)

Electrical Tunnel Personnel Hatch Area Pre-Action Valve Building Primary Auxiliary Building RHR and Containment Spray Equipment Vault Service Water Pumphouse, including Electrical Room Waste Processing Building, external walls and roof below Elev. 53'-O and including the following areas above Elev. 53'-O: Betw. Col. 1 & 2, Col's A to D to Elev. 86'-O Betw. Col. 4.9 & C, Col's B to E, to Elev. 86'-O Pipe Tunnel between Tank Farm and

PAB Safety-Related Electrical Manholes

  • Except where noted, all structures completely enclose the equipment housed therein. Refer to Reference 20 for an evaluation of a missile entering a safety-related structure which resulted in a probability of about 10

-6 per year.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.5-13 Revision:

Sheet: 8 1 of 1 TABLE 3.5-13 A IRCRAFT A CCIDENT PROBABILITY Type Aircraft N Area(mi)2 R Acc/Hr Exp Hr/(mi) 2 P(yr)-1 FB-111A 2,340 .0029 2.41x10-5 4.61x10-4 7.54x10-8 FB-111A* 12 .005 2.41x10-5 4.61x10-4 6.67x10-10 KC-135 2,903 .005 2.48x10-6 3.89x10-4 1.40x10-8 C-130 384 .005 2.2x10-6 3.89x10-4 1.64x10-9 Other** 459 .005 5.0x10-6 3.89x10-4 4.46x10-9 Commercial 46,720 .005 3.89x10-7 1.04x10-5 9.45x10-10 Total Aircraft Crash Probability = 9.71x10

-8

  • Adjustment for FB-111A that weigh more than 81,800 pounds. **Other includes C-123, F4, C5A, P3, B57, B52, and C-141 aircraft. It also includes an additional 200 aircraft that may overfly the area (Subsection 2.2.2.5).

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-1 Revision:

Sheet: 8 1 of 1 TABLE 3.6(B)-1 E SSENTIAL C OMPONENTS LOCATED O UTSIDE CONTAINMENT POTENTIALLY S USCEPTIBLE T O E FFECTS O F PIPING F AILURE Component Component Number Location RH-P-8A Primary Auxiliary Building Residual Heat Removal Pumps RH-P-8B Primary Auxiliary Building SI-P-6A Primary Auxiliary Building Safety Injection Pumps SI-P-6B Primary Auxiliary Building CS-P-2A Primary Auxiliary Building Chemical and Volume Control Charging Pumps CS-P-2B Primary Auxiliary Building CC-P-11A Primary Auxiliary Building CC-P-11B Primary Auxiliary Building CC-P-11C Primary Auxiliary Building Primary Component Cooling Water Pumps CC-P-11D Primary Auxiliary Building FW-P-37A Emergency Feedwater Pump-house Emergency Feedwater Pumps FW-P-37B Emergency Feedwater Pump- house Condensate Storage Tank CO-TK-25 Yard S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 1 of 18 TABLE 3.6(B)-2 HIGH E NERGY L INES Line No. Safety Class Essential Function Size Yes No P&ID 1-1 1 Primary Loop 29 X RC-20841 1-2 1 Primary Loop 31 X RC-20841 2-1 1 Primary Loop 31 X RC-20841 3-1 1 Primary Loop 271/2 X RC-20842 4-1 1 Primary Loop 29 X RC-20842 4-2 1 Primary Loop 31 X RC-20842 5-1 1 Primary Loop 31 X RC-20842 6-1 1 Primary Loop 271/2 X RC-20842 7-1 1 Primary Loop 29 X RC-20843 7-2 1 Primary Loop 31 X RC-20843 8-1 1 Primary Loop 31 X RC-20843 9-1 1 Primary Loop 271/2 X RC-20843 10-1 1 Primary Loop 29 X RC-20844 10-2 1 Primary Loop 31 X RC-20844 11-1 1 Primary Loop 31 X RC-20844 12-1 1 Primary Loop 271/2 X RC-20844 13-1 1 Reactor Coolant 12 X RC-20841 21-1 1 Reactor Coolant 4 X RC-20841 48-1 1 Reactor Coolant 4 X RC-20843, 20846 48-2 1 Reactor Coolant 6 X RC-20846 48-3 1 Reactor Coolant 4 X RC-20846 49-1 1 Reactor Coolant 14 X RC-20843, 20846 58-1 1 Reactor Coolant 12 X RC-20844 74-1 1 Reactor Coolant 6 X RC-20846 75-1 1 Reactor Coolant 6 X RC-20846 76-1 1 Reactor Coolant 6 X RC-20846 80-1 1 Reactor Coolant 6 X RC-20846 80-2 1 Reactor Coolant 3 X RC-20846 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 2 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 80-6 1 Reactor Coolant 3 X RC-20846 80-15 1 Reactor Coolant 6 X RC-20846 93-1 1 Reactor Coolant 2 X RC-20841 94-1 1 Reactor Coolant 2 X RC-20842 96-1 1 Reactor Coolant 2 X RC-20843 97-1 1 Reactor Coolant 3 X RC-20843 97-2 2 Reactor Coolant 3 X RC-20843, CS-20722 98-1 1 Reactor Coolant 2 X RC-20844 155-5 1 Residual Heat Removal 6 X RH-20662, SI-20450 155-17 1 Residual Heat Removal 10 X SI-20450 158-3 2 Residual Heat Removal 8 X RH-20663 158-4 2 Residual Heat Removal 6 X RH-20663 158-5 1 Residual Heat Removal 6 X RH-20663, SI-20450 158-17 1 Residual Heat Removal 10 X SI-20450 160-6 1 Residual Heat Removal 6 X RH-20663, RC-20844 160-17 1 Residual Heat 12 X RC-20844 162-2 1 Residual Heat Removal 6 X RH-20662, SI-20450 162-5 1 Residual Heat Removal 10 X SI-20450 163-1 2 Residual Heat Removal 6 X RH-20663 163-2 1 Residual Heat Removal 6 X RH-20663, SI-20450 163-4 1 Residual Heat Removal 10 X SI-20450 Removal 163-5 2 Residual Heat Removal 6 X RH-20663 177-1 2 Residual Heat Removal 2 X RH-20662, CS-20722 180-2 1 Residual Heat Removal 8 X RH-20663 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 3 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 180-3 1 Residual Heat Removal 6 X RH-20663, RC-20841 180-5 1 Residual Heat Removal 12 X RC-20841 201-1 2 Safety Injection 10 X SI-20450 201-2 1 Safety Injection 10 X SI-20450, RC-20841 202-1 2 Safety Injection 10 X SI-20450 202-2 1 Safety Injection 10 X SI-20450, RC-20841 203-1 2 Safety Injection 10 X SI-20450 203-2 1 Safety Injection 10 X SI-20450, RC-20843 204-1 2 Safety Injection 10 X SI-20450 204-2 1 Safety Injection 10 X SI-20450, RC-20844 232-1 2 Safety Injection 2 X SI-20450 232-2 2 Safety Injection 2 X SI-20450 234-1 2 Safety Injection 2 X SI-20450 234-2 2 Safety Injection 2 X SI-20450 236-1 2 Safety Injection 2 X SI-20450 236-2 2 Safety Injection 2 X SI-20450 238-1 2 Safety Injection 2 X SI-20450 238-2 2 Safety Injection 2 X SI-20450 240-1 2 Safety Injection 2 X SI-20450 240-2 2 Safety Injection 2 X SI-20450 242-1 2 Safety Injection 2 X SI-20450 242-2 2 Safety Injection 2 X SI-20450 244-1 2 Safety Injection 2 X SI-20450 244-2 2 Safety Injection 2 X SI-20450 246-1 2 Safety Injection 2 X SI-20450 246-2 2 Safety Injection 2 X SI-20450 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 4 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 251-3 2 Safety Injection 4 X SI-20446 251-5 1 Safety Injection 2 X SI-20446 251-6 1 Safety Injection 3 X SI-20446 251-7 1 Safety Injection 6 X SI-20446, RC-20843 256-3 2 Safety Injection 2 X SI-20446 256-4 1 Safety Injection 2 X SI-20446, RH-20663 258-1 2 Safety Injection 2 X SI-20446 258-2 1 Safety Injection 2 X SI-20446, RH-20663 259-3 1 Safety Injection 2 X SI-20446, RH-20662 260-2 1 Safety Injection 2 X SI-20446, RH-20662 261-2 1 Safety Injection 2 X SI-20448 261-3 1 Safety Injection 3 X SI-20448 261-4 1 Safety Injection 6 X SI-20446, RC-20842 270-2 1 Safety Injection 2 X SI-20446, RH-20663 272-2 2 Safety Injection 4 X SI-20447 272-3 2 Safety Injection 3 X SI-20447 272-4 1 Safety Injection 3 X SI-20447 272-5 1 Safety Injection 11/2 X SI-20447, RC-20841 272-9 2 Safety Injection 4 X SI-20447 273-1 1 Safety Injection 11/2 X SI-20447 273-5 1 Safety Injection 3 X SI-20447 274-1 1 Safety Injection 11/2 X SI-20447, RC-20844 274-5 1 Safety Injection 3 X SI-20447 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 5 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 275-4 1 Safety Injection 11/2 X SI-20447 275-6 1 Safety Injection 3 X SI-20447 318-1 2 Chem & Vol Control 3 X CS-20722 318-2 2 Chem & Vol Control 2 X CS-20722 318-9 2 Chem & Vol Control 2 X CS-20722 324-1 2 Chem & Vol Control 3 X CS-20722, CS-20726 325-1 2 Chem & Vol Control 2 X CS-20726 326-1 2 Chem & Vol Control 2 X CS-20726 327-1 2 Chem & Vol Control 3 X CS-20726 327-2 2 Chem & Vol Control 2 X CS-20726 328-1 2 Chem & Vol Control 2 X CS-20726 328-2 2 Chem & Vol Control 3 X CS-20726 328-3 2 Chem & Vol Control 2 X CS-20726 328-6 1 Chem & Vol Control 2 X CS-20726 328-7 1 Chem & Vol Control 11/2 X CS-20726 329-1 2 Chem & Vol Control 2 X CS-20726 329-4 2 Chem & Vol Control 2 X CS-20726 329-5 1 Chem & Vol Control 11/2 X CS-20726 330-1 2 Chem & Vol Control 2 X CS-20726 330-4 1 Chem & Vol Control 2 X CS-20726 330-5 1 Chem & Vol Control 11/2 X CS-20726 331-1 2 Chem & Vol Control 2 X CS-20726 331-4 1 Chem & Vol Control 2 X CS-20726 331-5 1 Chem & Vol Control 11/2 X CS-20726 348-1 NNS Chem & Vol Control 3 X CS-20724 354-1 NNS Chem & Vol Control 3 X CS-20724 355-1 2 Chem & Vol Control 3 X CS-20722, CS-20725 355-6 2 Chem & Vol Control 3 X CS-20722 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 6 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 356-1 2 Chem & Vol Control 4 X CS-20725 356-2 2 Chem & Vol Control 3 X CS-20725 358-1 2 Chem & Vol Control 2 X CS-20725 358-2 2 Chem & Vol Control 3 X CS-20725 358-3 2 Chem & Vol Control 2 X CS-20725 360-1 2 Chem & Vol Control 3 X CS-20722 360-2 2 Chem & Vol Control 2 X CS-20722 360-3 2 Chem & Vol Control 2 X CS-20722 360-4 2 Chem & Vol Control 3 X CS-20722 360-5 2 Chem & Vol Control 2 X CS-20722 360-6 2 Chem & Vol Control 2 X CS-20722 360-7 2 Chem & Vol Control 3 X CS-20722 360-8 2 Chem & Vol Control 4 X CS-20722 360-9 2 Chem & Vol Control 3 X CS-20722 360-25 2 Chem & Vol Control 3 X CS-20722 360-26 2 Chem & Vol Control 3 X CS-20722 361-1 2 Chem & Vol Control 2 X CS-20722 362-1 2 Chem & Vol Control 4 X CS-20725 363-1 2 Chem & Vol Control 2 X CS-20725 363-2 2 Chem & Vol Control 3 X CS-20725 363-3 2 Chem & Vol Control 2 X CS-20725 364-1 2 Chem & Vol Control 4 X CS-20725 364-2 2 Chem & Vol Control 3 X CS-20725 364-9 2 Chem & Vol Control 11/2 X CS-20725 365-1 2 Chem & Vol Control 2 X CS-20725 365-2 1 Chem & Vol Control 2 X CS-20722, RC-20846 366-1 2 Chem & Vol Control 3 X CS-20722 366-2 1 Chem & Vol Control 3 X CS-20722, RC-20841 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 7 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 367-1 2 Chem & Vol Control 2 X CS-20725 368-1 2 Chem & Vol Control 3 X CS-20722 368-2 1 Chem & Vol Control 3 X CS-20722 374-1 2 Chem & Vol Control 4 X CS-20725 376-1 NNS Chem & Vol Control 3 X CS-20724 431-1 2 Chem & Vol Control 3 X CS-20727, CS-20722 431-2 2 Chem & Vol Control 4 X CS-20727 432-1 2 Chem & Vol Control 4 X CS-20727 432-2 2 Chem & Vol Control 3 X CS-20727, 511-9 3 Primary Component Cooling 3/4 X CS-20722 525-1 NNS Chem & Vol Control 4 X CS-20724 525-2 NNS Chem & Vol Control 3 X CS-20724 525-3 NNS Chem & Vol Control 3 X CS-20724 526-1 NNS Chem & Vol Control 2 X CS-20724 526-2 NNS Chem & Vol Control 2 X CS-20724 526-8 NNS Chem & Vol Control 11/2 X CS-20724 526-9 NNS Chem & Vol Control 11/2 X CS-20724 526-10 NNS Chem & Vol Control 3 X CS-20724 526-13 NNS Chem & Vol Control 21/2 X CS-20724 534-1 NNS Chem & Vol Control 3 X CS-20724 541-3 3 Primary Component Cooling 3/4 X CS-20722 541-5 3 Primary Component Cooling 3/4 X CS-20722 816-1 3 Primary Component Cooling 6 X CS-20206 816-5 3 Primary Component Cooling 8 X CS-20206 816-6 3 Primary Component Cooling 2 X CS-20206 816-10 3 Primary Component Cooling 3 X CS-20206 816-11 3 Primary Component Cooling 6 X CS-20206 816-12 3 Primary Component Cooling 3/4 X CS-20206 1301-1 2 Steam Gen Blowdown 2 X RC-20841 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 8 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 1301-2 2 Steam Gen Blowdown 3 X RC-20841, SB-20626 1301-3 2 Steam Gen Blowdown 2 X SB-20626 1301-4 2 Steam Gen Blowdown 3 X SB-20626 1301-5 NNS Steam Gen Blowdown 3 X SB-20626 1301-16 NNS Steam Gen Blowdown 3 X SB-20626 1301-22 NNS Steam Gen Blowdown 11/2 X SB-20626 1301-23 NNS Steam Gen Blowdown 4 X SB-20626 1303-1 2 Steam Gen Blowdown 2 X RC-20841 1304-1 2 Steam Gen Blowdown 2 X RC-20842 1304-2 2 Steam Gen Blowdown 3 X RC-20842, SB-20626 1304-3 2 Steam Gen Blowdown 2 X SB-20626 1304-4 2 Steam Gen Blowdown 3 X SB-20626 1304-5 NNS Steam Gen Blowdown 3 X SB-20626 1304-17 NNS Steam Gen Blowdown 3 X SB-20626 1304-24 NNS Steam Gen Blowdown 11/2 X SB-20626 1304-25 NNS Steam Gen Blowdown 4 X SB-20626 1306-2 2 Steam Gen Blowdown 2 X RC-20842 1307-1 2 Steam Gen Blowdown 2 X RC-20843 1307-2 2 Steam Gen Blowdown 3 X RC-20843, SB-20626 1307-3 2 Steam Gen Blowdown 2 X SB-20626 1307-4 2 Steam Gen Blowdown 3 X SB-20626 1307-5 NNS Steam Gen Blowdown 3 X SB-20626 1307-17 NNS Steam Gen Blowdown 3 X SB-20626 1307-22 NNS Steam Gen Blowdown 11/2 X SB-20626 1307-23 NNS Steam Gen Blowdown 4 X SB-20626 1309-2 2 Steam Gen Blowdown 2 X RC-20843 1310-1 2 Steam Gen Blowdown 2 X RC-20844 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 9 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 1310-2 2 Steam Gen Blowdown 3 X RC-20844, SB-20626 1310-3 2 Steam Gen Blowdown 2 X SB-20626 1310-4 2 Steam Gen Blowdown 3 X SB-20626 1310-5 NNS Steam Gen Blowdown 3 X SB-20626 1310-16 NNS Steam Gen Blowdown 3 X SB-20626 1310-22 NNS Steam Gen Blowdown 11/2 X SB-20626 1310-23 NNS Steam Gen Blowdown 4 X SB-20626 1312-2 2 Steam Gen Blowdown 2 X RC-20844 1317-1 NNS Steam Gen Blowdown 10 X SB-20626 1317-7 NNS Steam Gen Blowdown 14 X SB-20626 1319-1 NNS Steam Gen Blowdown 10 X SB-20626 1320-1 NNS Steam Gen Blowdown 8 X SB-20626 1320-2 NNS Steam Gen Blowdown 3 X SB-20626 1320-8 NNS Steam Gen Blowdown 4 X SB-20626 1321-1 NNS Steam Gen Blowdown 3 X SB-20626 1321-4 NNS Steam Gen Blowdown 4 X SB-20626 1350-3 NNS Steam Gen Blowdown 10 X SB-20626 2302-2 NNS Aux Steam 8 X AS-20570, AS-20569 2302-5 NNS Aux Steam 4 X AS-20570 2302-8 NNS Aux Steam 4 X AS-20570 2302-14 NNS Aux Steam 2 X AS-20571 2302-16 NNS Aux Steam 2 X AS-20571 2302-19 NNS Aux Steam 4 X AS-20570 2302-30 NNS Aux Steam 4 X AS-20571 2302-32 NNS Aux Steam 8 X AS-20570, AS-20571 2303-1 NNS Aux Steam 6 X AS-20570 2303-2 NNS Aux Steam 4 X AS-20570 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 10 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 2303-3 NNS Aux Steam 4 X AS-20570 2303-5 NNS Aux Steam 2 X AS-20570 2303-6 NNS Aux Steam 3 X AS-20570 2304-1 NNS Aux Steam 3 X AS-20570 2304-2 NNS Aux Steam 11/2 X AS-20570 2304-3 NNS Aux Steam 3 X AS-20570 2306-1 NNS Aux Steam 8 X AS-20570 2306-2 NNS Aux Steam 4 X AS-20570 2306-4 NNS Aux Steam 8 X AS-20570 2306-5 NNS Aux Steam 10 X AS-20570 2309-1 NNS Aux Steam 2 X AS-20571 2309-2 NNS Aux Steam 2 X AS-20571 2339-1 NNS Aux Steam 11/2 X AS-20571 2339-2 NNS Aux Steam 11/2 X AS-20571 2341-1 NNS Aux Steam 11/2 X AS-20571 2341-4 NNS Aux Steam 2 X AS-20571 2341-5 NNS Aux Steam 11/2 X AS-20571 2364-1 NNS Aux Steam 3 X AS-20570 2365-6 NNS Aux Steam 2 X AS-20570 2366-1 NNS Aux Steam 2 X AS-20570 2401-1 NNS Aux Steam Condensate 21/2 X ASC-20906 2401-2 NNS Aux Steam Condensate 3 X ASC-20906 2401-3 NNS Aux Steam Condensate 21/2 X ASC-20906 2401-4 NNS Aux Steam Condensate 11/2 X ASC-20906 2402-2 NNS Aux Steam Condensate 2 X ASC-20906 2402-3 NNS Aux Steam Condensate 3 X ASC-20906 2403-1 NNS Aux Steam Condensate 11/2 X ASC-20906 2404-1 NNS Aux Steam Condensate 11/2 X ASC-20906 2404-2 NNS Aux Steam Condensate 3 X ASC-20906 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 11 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 2404-3 NNS Aux Steam Condensate 3 X ASC-20906 2404-4 NNS Aux Steam Condensate 6 X ASC-20906 2404-5 NNS Aux Steam Condensate 3 X ASC-20906 2404-6 NNS Aux Steam Condensate 11/2 X ASC-20906 2404-8 NNS Aux Steam Condensate 3 X ASC-20906 2406-1 NNS Aux Steam Condensate 4 X ASC-20906, ASC-20907 2406-2 NNS Aux Steam Condensate 6 X ASC-20906 2406-3 NNS Aux Steam Condensate 4 X ASC-20906 2406-4 NNS Aux Steam Condensate 3 X ASC-20906 2406-5 NNS Aux Steam Condensate 3 X ASC-20906 2407-4 NNS Aux Steam Condensate 11/2 X ASC-20907 2409-4 NNS Aux Steam Condensate 11/2 X ASC-20907 2410-1 NNS Aux Steam Condensate 2 X ASC-20906 2410-4 NNS Aux Steam Condensate 11/2 X ASC-20906 2433-1 NNS Aux Steam Condensate 2 X ASC-20906 2433-2 NNS Aux Steam Condensate 11/4 X ASC-20906 2433-3 NNS Aux Steam Condensate 11/4 X ASC-20906 2433-4 NNS Aux Steam Condensate 11/4 X ASC-20906 2437-1 NNS Aux Steam Condensate 21/2 X ASC-20906 2437-2 NNS Aux Steam Condensate 3 X ASC-20906 2437-3 NNS Aux Steam Condensate 21/2 X ASC-20906 2437-4 NNS Aux Steam Condensate 11/2 X ASC-20906 2438-2 NNS Aux Steam Condensate 2 X ASC-20906 2439-1 NNS Aux Steam Condensate 11/2 X ASC-20906 2439-2 NNS Aux Steam Condensate 11/2 X ASC-20906 2439-7 NNS Aux Steam Condensate 11/2 X ASC-20907 2440-1 NNS Aux Steam Condensate 11/2 X ASC-20906 2441-1 NNS Aux Steam Condensate 11/2 X ASC-20907 2441-2 NNS Aux Steam Condensate 11/2 X ASC-20907 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 12 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 2441-4 NNS Aux Steam Condensate 2 X ASC-20907 2441-7 NNS Aux Steam Condensate 11/2 X ASC-20907 2442-1 NNS Aux Steam Condensate 2 X ASC-20906 2442-2 NNS Aux Steam Condensate 2 X ASC-20906 2450-1 NNS Aux Steam Condensate 11/4 X ASC-20906 2450-2 NNS Aux Steam Condensate 11/2 X ASC-20906 2451-1 NNS Aux Steam Condensate 11/2 X ASC-20906 4000-1 2 Mainsteam 32 X MS-20580, MS-20583 4000-2 2 Mainsteam 30 X MS-20580 4000-3 NNS Mainsteam 30 X MS-20583 4000-8 2 Mainsteam 6 X MS-20580 4000-11 2 Mainsteam 6 X MS-20580 4000-32 2 Mainsteam 6 X MS-20580 4000-34 2 Mainsteam 6 X MS-20580 4000-40 2 Mainsteam 30 X MS-20580 4000-41 2 Mainsteam 30 X MS-20580, MS-20583 4000-48 2 Mainsteam 2 X MS-20580 4001-1 2 Mainsteam 32 X MS-20581 4001-2 2 Mainsteam 30 X MS-20581 4001-3 NNS Mainsteam 30 X MS-20583 4001-8 2 Mainsteam 6 X MS-20581 4001-11 2 Mainsteam 6 X MS-20581 4001-32 2 Mainsteam 6 X MS-20581 4001-34 2 Mainsteam 6 X MS-20581 4001-39 2 Mainsteam 30 X MS-20581 4001-41 2 Mainsteam 30 X MS-20580, MS-20583 4001-46 2 Mainsteam 2 X MS-20581 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 13 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 4002-1 2 Mainsteam 32 X MS-20581 4002-2 2 Mainsteam 30 X MS-20581 4002-3 NNS Mainsteam 30 X MS-20583 4002-9 2 Mainsteam 6 X MS-20581 4002-30 2 Mainsteam 6 X MS-20581 4002-32 2 Mainsteam 6 X MS-20581 4002-36 2 Mainsteam 30 X MS-20581 4002-37 2 Mainsteam 30 X MS-20581, MS-20583 4002-41 2 Mainsteam 2 X MS-20581 4003-1 2 Mainsteam 32 X MS-20580 4003-2 2 Mainsteam 30 X MS-20580 4003-3 NNS Mainsteam 30 X MS-20580 4003-9 2 Mainsteam 6 X MS-20580 4003-30 2 Mainsteam 6 X MS-20580 4003-32 2 Mainsteam 6 X MS-20580 4003-36 2 Mainsteam 30 X MS-20580 4003-37 2 Mainsteam 30 X MS-20580, MS-20583 4003-41 2 Mainsteam 2 X MS-20580 4004-1 NNS Mainsteam 48 X MS-20583 4005-1 NNS Mainsteam 24 X MS-20583, MS-20585 4006-1 NNS Mainsteam 30 X MS-20583 4006-2 NNS Mainsteam 6 X MS-20583 4007-1 NNS Mainsteam 30 X MS-20583 4007-2 NNS Mainsteam 6 X MS-20583 4008-1 NNS Mainsteam 30 X MS-20583 4008-2 NNS Mainsteam 6 X MS-20583 4009-1 NNS Mainsteam 30 X MS-20583 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 14 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 4009-2 NNS Mainsteam 6 X MS-20583 4010-1 NNS Mainsteam 24 X MS-20583, MS-20585 4010-33 NNS Mainsteam 6 X MS-20583 4366-1 3 Diesel Generator - Air 2 X DG-20460 4367-1 3 Diesel Generator - Air 2 X DG-20460 4368-1 3 Diesel Generator - Air 2 X DG-20465 4369-1 3 Diesel Generator - Air 2 X DG-20465 4454-1 NNS Mainsteam 2 X MS-20587 4510-3 NNS Mainsteam Drain 2 X MS-20587 4511-3 NNS Mainsteam Drain 2 X MS-20587 4511-5 NSS Mainsteam Drain 2 X MS-20587 4513-3 NNS Mainsteam Drain 2 X MS-20587 4513-4 NNS Mainsteam Drain 2 X MS-20583 4515-3 NNS Mainsteam Drain 2 X MS-20587 4515-4 NSS Mainsteam Drain 2 X MS-20583 4517-3 NNS Mainsteam Drain 2 X MS-20587 4517-4 NNS Mainsteam Drain 2 X MS-20583 4519-3 NNS Mainsteam Drain 2 X MS-20587 4519-4 NNS Mainsteam Drain 2 X MS-20583 4606-2 NNS Feedwater 18 X FW-20686 4606-3 2 Feedwater 18 X FW-20686 4606-4 2 Feedwater 16 X FW-20686 4606-12 2 Feedwater 2 X FW-20686 4606-15 2 Feedwater 4 X FW-20686, FW-20688 4607-2 NNS Feedwater 18 X FW-20686 4607-3 2 Feedwater 18 X FW-20686 4607-4 2 Feedwater 16 X FW-20686 4607-12 2 Feedwater 2 X FW-20686 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 15 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 4607-15 2 Feedwater 4 X FW-20686, FW-20688 4608-2 NNS Feedwater 18 X FW-20686 4608-3 2 Feedwater 18 X FW-20686 4608-4 2 Feedwater 16 X FW-20686 4608-12 2 Feedwater 2 X FW-20686 4608-15 2 Feedwater 4 X FW-20686, FW-20688 4609-2 NNS Feedwater 18 X FW-20686 4609-3 2 Feedwater 18 X FW-20686 4609-4 2 Feedwater 16 X FW-20686 4609-12 2 Feedwater 2 X FW-20686 4609-15 2 Feedwater 4 x FW-20686, FW-20688 4614-2 2 Feedwater 4 X FW-20686 4615-2 2 Feedwater 4 X FW-20686 4616-2 2 Feedwater 4 X FW-20686 4617-2 2 Feedwater 4 X FW-20686 5198-1 NNS Aux Steam 12 X AS-20569 5198-2 NNS Aux Steam 2 X AS-20569 5198-3 NNS Aux Steam 2 X AS-20569 5198-8 NNS Aux Steam 12 X AS-20569 5198-9 NNS Aux Steam 2 X AS-20569 5198-11 NNS Aux Steam 2 X AS-20569 5198-13 NNS Aux Steam 11/2 X AS-20569 5198-18 3 Aux Steam 12 X AS-20569 5198-20 3 Aux Steam 12 X AS-20569 5198-21 NNS Aux Steam 11/2 X AS-20569 5198-22 NNS Aux Steam 12 X AS-20569 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 16 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 5230-1 NNS Aux Steam 6 X ASC-20908, ASC-20909, ASC-20906 5231-1 NNS Aux Steam Condensate 2 X ASC-20908 8751-1 NNS Aux Steam Heat 11/4 X ASC-20908 8757-1 NNS Aux Steam Condensate 3 X ASC-20906 8757-2 NNS Aux Steam Condensate 2 X ASC-20906 8757-5 NNS Aux Steam Heat 21/2 X ASC-20906 8767-1 NNS Aux Steam Heat 11/4 X ASC-20908 8863-1 NNS Aux Steam Condensate 2 X ASC-20906 8864-1 NNS Aux Steam Condensate 21/2 X ASC-20906 8864-2 NNS Aux Steam Condensate 2 X ASC-20906 9000-1 NNS Hot Water Supply 6 X HW-20051 9001-1 NNS Hot Water Supply 6 X HW-20051 9001-2 NNS Hot Water Supply 3 X HW-20051 9002-1 NNS Hot Water Supply 4 X HW-20051 9003-1 NNS Hot Water Supply 4 X HW-20051 9004-1 NNS Hot Water Supply 11/4 X HW-20051 9006-1 NNS Hot Water Supply 2 X HW-20051 9007-1 NNS Hot Water Supply 4 X HW-20051 9008-1 NNS Hot Water Supply 4 X HW-20051 9009-1 NNS Hot Water Supply 11/4 X HW-20051 9011-1 NNS Hot Water Supply 2 X HW-20051 9012-1 NNS Hot Water Supply 11/2 X HW-20051 9013-1 NNS Hot Water Supply 2 X HW-20051 9014-1 NNS Hot Water Supply 11/2 X HW-20051 9015-1 NNS Hot Water Supply 2 X HW-20051 9022-1 NNS Hot Water Supply 4 X HW-20051 9023-1 NNS Hot Water Supply 4 X HW-20051 9024-1 NNS Hot Water Supply 4 X HW-20051 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 17 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 9025-1 NNS Hot Water Supply 4 X HW-20051 9026-1 NNS Hot Water Supply 4 X HW-20051 9027-1 NNS Hot Water Supply 4 X HW-20051 9030-1 NNS Hot Water Supply 2 X HW-20056 9030-2 NNS Hot Water Supply 11/4 X HW-20056 9042-1 NNS Hot Water Supply 2 X HW-20051 9043-1 NNS Hot Water Supply 2 X HW-20051 9044-1 NNS Hot Water Supply 2 X HW-20051 9045-1 NNS Hot Water Supply 2 X HW-20051 9050-1 NNS Hot Water Supply 2 X HW-20051 9051-1 NNS Hot Water Supply 2 X HW-20051 9052-1 NNS Hot Water Supply 2 X HW-20051 9053-1 NNS Hot Water Supply 2 X HW-20051 9054-1 NNS Hot Water Supply 2 X HW-20051 9072-1 NNS Hot Water Supply 2 X HW-20051 9200-1 NNS Hot Water Return 6 X HWR-20051 9201-1 NNS Hot Water Return 4 X HWR-20051 9202-1 NNS Hot Water Return 4 X HWR-20051 9208-1 NNS Hot Water Return 4 X HWR-20051 9209-1 NNS Hot Water Return 2 X HWR-20051 9210-1 NNS Hot Water Return 2 X HWR-20051 9211-1 NNS Hot Water Return 2 X HWR-20051 9212-1 NNS Hot Water Return 2 X HWR-20051 9213-1 NNS Hot Water Return 2 X HWR-20051 9214-1 NNS Hot Water Return 4 X HWR-20051 9215-1 NNS Hot Water Return 2 X HWR-20051 9216-1 NNS Hot Water Return 2 X HWR-20051 9217-1 NNS Hot Water Return 2 X HWR-20051 9218-1 NNS Hot Water Return 2 X HWR-20051 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.6(B)-2 Revision:

Sheet: 10 18 of 18 Line No. Safety Class Essential Function Size Yes No P&ID 9219-1 NNS Hot Water Return 2 X HWR-20051 9220-2 NNS Hot Water Return 11/4 X HWR-20056 9220-3 NNS Hot Water Return 2 X HWR-20051, HWR-20056 9221-3 NNS Hot Water Return 11/2 X HWR-20056 9222-3 NNS Hot Water Return 11/2 X HWR-20056 9223-2 NNS Hot Water Return 11/4 X HWR-20056 9223-3 NNS Hot Water Return 11/2 X HWR-20056 9268-3 NNS Hot Water Return 11/2 X HW-20056 9270-2 NNS Hot Water Return 11/2 X HW-20056 9270-3 NNS Hot Water Return 11/4 X HW-20056 9271-2 NNS Hot Water Return 11/2 X HW-20052 9271-6 NNS Hot Water Return 11/2 X HW-20052 9272-3 NNS Hot Water Return 11/2 X HW-20052 9273-3 NNS Hot Water Return 11/2 X HW-20052 9274-3 NNS Hot Water Return 11/2 X HW-20052 9275-3 NNS Hot Water Return 11/2 X HW-20052 9826-1 NNS Hot Water Supply 11/2 X HW-20056 9830-1 NNS Hot Water Supply 11/2 X HW-20056 9831-1 NNS Hot Water Supply 11/2 X HW-20056 9832-1 NNS Hot Water Supply 11/2 X HW-20056 9833-1 NNS Hot Water Supply 11/2 X HW-20056 9834-3 NNS Hot Water Supply 11/2 X HW-20056 9835-1 NNS Hot Water Supply 11/2 X HW-20056 9835-4 NNS Hot Water Supply 11/2 X HW-20056 9836-3 NNS Hot Water Supply 11/2 X HW-20056 9849-1 NNS Hot Water Supply 11/2 X HW-20052 9849-7 NNS Hot Water Supply 11/2 X HW-20052 9850-3 NNS Hot Water Supply 11/2 X HW-20052 9851-3 NNS Hot Water Supply 11/2 X HW-20052 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.6(N)-1 Revision:

Sheet: 8 1 of 1 TABLE 3.6(N)-1* POSTULATED B REAK L OCATIONS F OR THE L OCA A NALYSIS O F THE P RIMARY COOLANT L OOP Location of Postulated Rupture Type Break Opening Area* 1. Residual Heat Removal (RHR) Line/Primary Coolant Loop Connection Guillotine (viewed from the RHR line) Cross-Sectional Flow Area of the RHR Line 2. Accumulator (ACC) Line/ Primary Coolant Loop Connection Guillotine (viewed from the ACC line) Cross-Sectional Flow Area of the ACC line 3. Pressurizer Surge (PS) Line/Primary Coolant Loop Connection Guillotine (viewed from the PS line) Cross-Sectional Flow Area of the PS line

  • Refer to Figure 3.6(N)-2 for location of postulated breaks in reactor coolant loop. ** Less break opening area will be used if justified by analysis, experiments or considerations of physical restraints such as concrete walls or structural steel.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-1 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-1 D AMPING V ALUES* (Percent of Critical Damping)

Structure or Component Operating Basis Earthquake Safe Shutdown Earthquake Equipment and large-diameter piping systems, pipe diameter greater than 12 in.

2 3 Small-diameter piping systems, diameter equal to or less than 12 in.

1 2 Welded steel structures 2 4 Bolted steel structures 4 7 Prestressed concrete structures 2 5 Reinforced concrete structures 4 7 Reference Newark, N.M., John A. Blume, and Kanwar K. Kapur, "Design Response Spectra for Nuclear Power Plants," ASCE Structural Engineering Meeting, San Francisco, April 1973.

  • For seismic piping analysis, the values in Figure 3.7(B)-38 may be used as an alternative.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-2 Revision:

Sheet: 8 1 of 2 TABLE 3.7(B)-2 CONTAINMENT SHELL - S EISMIC MODEL a. Beam Section Properties Beam No. Area (in 2 x10 10) Moment of Inertia (in 4 x10 10) Shear Shape*

Factor 1 2.272 1.462 0.53 2 2.272 3.342 0.53 3 2.272 5.402 0.53 4 2.272 7.035 0.53 5 2.272 7.926 0.53 6 2.272 8.371 0.53 7 2.942 11.067 0.53 8 2.942 11.067 0.53 9 2.942 11.067 0.53 10 2.942 11.067 0.53 11 2.942 11.067 0.53 12 2.942 11.067 0.53 13 2.942 11.067 0.53 14 2.942 11.067 0.53 15 2.942 11.067 0.53 16 2.942 11.067 0.53 17 2.942 11.067 0.53 18 2.942 11.067 0.53 19 2.942 11.067 0.53 20 2.942 11.067 0.53

  • Area x shear shape factor = Shear area

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-2 Revision:

Sheet: 8 2 of 2 b. Modal Weight & Weig ht Moment of Inertia Node W (Kipx10 3) (K-in 2 x10 8) 1 1.166 0.761 2 2.332 2.483 3 3.129 6.437 4 3.395 9.158 5 2.864 9.423 6 2.864 10.262 7 2.624 13.581 8 4.776 18.102 9 5.167 19.605 10 4.071 15.416 11 4.424 18.781 12 5.872 22.337 13 4.893 18.589 14 3.915 14.800 15 3.367 12.715 16 4.697 17.859 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-3 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-3 CONTAINMENT I NTERNALS M ATHEMATICAL MODEL a. Summary of Section Properties Beam No. Area (in 2) IX2 (in 4) IX3 (in 4) J (in 4) SF2* 0.33 SF3 0.70 30,31 3.523x10 5 0.171x10 10 3.406x10 10 5.818x10 10 0.33 0.70 27 3.598x10 5 0.348x10 10 0.517x10 10 6.556x10 10 0.40 0.62 25 4.093x10 5 0.380x10 10 0.549x10 10 6.407x10 10 0.41 0.61

  • SF2 = Shear shape factor for bending about X2(N-S) axis
b. Summary of Nodal Weights Node WX1 WX2,WX3 WX4 P4 (lbx10 6) (lbx10 6) (lb-in 2 x10 11) 29 6.061 9.711 23.581 26 6.913 8.727 24.187 23 6.780 6.780 11.694 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-4 Revision:

Sheet: 8 1 of 2 TABLE 3.7(B)-4 P RIMARY AUXILIARY B UILDING - S EISMIC MODEL a. Beam Properties Elev (ft) Center of Rigidity (in) Moment of Inertia (in 4) Shear Shape Factor*

Beam No. From To

Area (in 2) Xcr Ycr Ix-x Iy-y Torsional Constant, J (in 4) SFX SFY 2 108 81 61,632 1050.72 323.40 8.665x10 8 1.2825x10 9 6.3348x10 9 0.4673 0.5327 5 81 53 210,586 722.64 539.18 4.3234x10 9 1.9511x10 10 5.7646x10 10 0.4721 0.5279 8 53 25 194,256 757.20 483.12 4.0186x10 9 1.1467x10 10 6.7102x10 10 0.4694 0.5306 11 25 73 39,552 853.20 536.52 5.6858x10 9 1.7503x10 10 8.6490x10 10 0.6166 0.3834

  • Shear area for bending about X-X = Area X SFX Shear area for bending about Y-Y = Area X SFY

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-4 Revision:

Sheet: 8 2 of 2 b. Weight, Weight Moment of Inertia, Center of Weight Mass Center Weight Moment of Inertia, Mass Center (in) Node No. Elev (ft) Weight (10 6lbs) IX IY IZ XBAR YBAR 2 108 2.1897 100.166 116.580 213.590 1044.00 318.00 5 81 9.4550 860.672 2481.696 3326.308 811.58 453.26 8 53 11.0715 1039.794 3167.405 4185.158 741.79 501.62 11 25 11.2919 996.351 3068.816 4048.544 851.51 487.97 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-5 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-5 P RIMARY AUXILIARY B UILDING - S EISMIC A NALYSIS D YNAMIC P ROPERTIES

- COUPLED TORSION MODEL Participation Factor Mode No. Natural Frequency (Hz) Generated Weight, (10 7 lbs) X2 (N-S) X3 (E-W) Remarks - Type of Mode 1 5.0775 0.7331 0.0093 -1.7114 Bending About X2 (N-S) Axis 2 7.0437 0.8780 1.6649 -0.0155 Bending About X3 (E-W) Axis 3 13.0131 5.1185 -0.0639 -.0639 Torsion 4 16.8425 0.4718 0.0834 -1.0227 Bending About X2 (N-S) Axis 5 19.0848 0.4287 -1.0773 -0.0534 Bending About X3 (E-W) Axis 6 26.5067 1.1298 -0.0082 0.4578 Bending About X2 (N-S) Axis 7 27.7903 0.9912 0.5293 0.0008 Bending About X3 (E-W) Axis 8 32.3146 101.8944 0.0055 0.0033 Torsion 9 35.9770 2.1381 0.2764 0.0142 Bending About X3 (E-W) Axis 10 37.3225 1.3661 -0.0126 0.5899 Bending About X2 (N-S) Axis 11 37.7250 16.0945 0.0011 0.0138 Torsion 12 48.1893 32.5880 0.0052 0.0051 Torsion S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-6 Revision:

Sheet: 8 1 of 2 TABLE 3.7(B)-6 P RIMARY AUXILIARY B UILDING - V ERTICAL S EISMIC A NALYSIS MODEL a. Nodal Weights and Spring Constants El. (ft) Node No. Wt. (Kips) Frequency (Hz) Constant (10 3 Kips/in)

Mode Classification 108 1 1655 21.599 - 1st Structural Frequency 2 535 4.054 0.901 Local Frequency 81 3 7615 21.599 - 1st Structural Frequency 4 519 7.184 2.749 Local Frequency 5 964 10.225 10.449 Local Frequency 6 228 12.654 3.787 Local Frequency 7 129 24.220 7.403 Local Frequency 65.75 9 40 8.097 0.268 Local Frequency 53 10 9306 21.599 - 1st Structural Frequency 11 150 7.596 0.885 Local Frequency 12 1338 8.950 11.073 Local Frequency 13 168 10.173 1.786 Local Frequency 14 110 18.424 3.888 Local Frequency 25 15 9488 21.599 - 1st Structural Frequency 16 385 7.497 2.213 Local Frequency 17 740 8.885 5.989 Local Frequency 18 249 11.591 3.423 Local Frequency 19 430 17.032 12.846 Local Frequency 7 20 375 9.300 3.314 Local Frequency 21 96.0 14.401 2.034 Local Frequency S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-6 Revision:

Sheet: 8 2 of 2 Beam No. Axial Area (in

2) 1 61632 2 210586 3 210586 4 194256 5 339552 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-7 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-7 CONTROL A ND D IESEL GENERATOR B UILDING S EISMIC A NALYSIS MODEL a. Summary of Nodal Weights NODE NO. WEIGHT (lbs) I-(lb-in 2) About X1 (Vert.)

3 7.9198x10 6 2.5171x10 12 6 5.6269x10 6 1.1731x10 12 9 8.6589x10 6 2.6786x10 12 12 7.0052x10 6 1.3945x10 12 15 7.5876x10 6 2.6709x10 12 b. Summary of Beam Properties Area Moment of Inertia Beam No. Area (in 2) Torsional Constant (J) (in 4) About X2 IX2 (in 4) About X3 IX3 (in 4) SF2 SF3 1 48744 6.524x10 9 2.767x10 10 2.355x10 9 0.370 0.520 2 51840 7.387x10 9 2.767x10 10 2.355x10 10 0.511 0.489 3 64800 1.100x10 10 8.887x10 9 2.355x10 9 0.609 0.391 4 64800 1.100x10 10 2.767x10 10 2.355x10 9 0.609 0.391 5 64800 1.100x10 10 2.767x10 10 2.355x10 9 0.609 0.391 6 64800 1.100x10 10 8.887x10 9 2.355x10 10 0.609 0.391 7 116640 9.314x10 9 2.767x10 10 4.710x10 9 0.565 0.435 8 101448 1.293x10 10 2.767x10 10 3.188x10 8 0.567 0.437 9 116640 9.320x10 9 2.767x10 10 4.710x10 9 0.565 0.435 10 116640 1.293x10 10 2.767x10 10 4.710x10 9 0.565 0.435 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-8 Revision:

Sheet: 8 1 of 4 TABLE 3.7(B)-8 F UEL STORAGE B UILDING S EISMIC A NALYSIS MODEL a. Element (Beam) Properties and Nodal Weights Moment of Inertia (10 10 in 4) Shear Shape Factor For Bending Model Beam No. Area (in 2) Torsional Constant (J)

(10 10 in 4) About X2 About X3 SF2 SF3 1 61776 2.111 .4066 .3203 .647 .523 2 75456 1.940 .4066 .4805 .530 .491 3 61776 1.976 .4066 .3203 .568 .435 4 75456 1.940 .4066 .4805 .465 .491 5 61776 2.428 .4066 .3203 .590 .364 6 75456 1.985 .4066 .4805 .483 .573 7 61776 2.428 .4066 .3203 .590 .364 8 75456 1.985 .4066 .4805 .483 .573 Model for N - S Motion (X2) 9 378300 6.849 1.186 1.171 .522 .634 b. Nodal Weight and We ight Moment of Inertia

Node No. Translational Inertia (lbs) X2 Rotational Inertia (10 11 lb-in 2) About X1 2 5051000 14.21 4 3216000 9.953 5 797000 .846 6 941000 2.709 7 1154000 2.493 8 648000 1.805 9 873000 2.047 12 1887000 5.683 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-8 Revision:

Sheet: 8 2 of 4 c. Element (Beam) Properties and Nodal Weights Moment of Inertia (10 10 in 4) Shear Shape Factor For Bending Model Beam No. Area (in 2) Torsional Constant (J) 10 10 in 4 About X2 About X3 SF 2 SF 3 1 137232 4.051 8.132 8.008 .583 .486 2 137232 3.916 8.l32 8.008 .512 .422 3 137232 4.232 8.l32 8.008 .537 .484 4 137232 4.232 8.l32 8.008 .537 .484 5 378300 6.849 1.186 1.171 .522 .634 6 15329 - .0003105 - - - Model for E - W (X3) 7 15329 - .0003105 - - - d. Nodal Weights and Weights Moment of Inertia

Node No. Translational Inertia (lbs) X3 Rotational Inertia (10 12 lb-in 2) About X1 2 4903000 1.379 5 3707000 1.270 6 843000 - 8 1859000 0.894 9 1344000 0.644 12 1772000 0.568 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-8 Revision:

Sheet: 8 3 of 4 e. Nodal Weight and Beam Properties (Vertical Model)

Node No. Nodal Weight (lbs) Element No. Axial Area (in 2) Spring Constant (lbs/in) 1 850000 1 137232 1389000 2 1203000 2 137232 3718000 3 271000 3 137232 5049900 4 2727000 4 137232 - 5 563000 5 378300 920300 6 140000 - - 1033000 7 3310000 - - - 8 2093000 - - - 9 1521000 - - - 10 1656000 - - - 11 178000 - - 1473000 12 84000 - - 1236000 13 430000 - - 8610000 14 90000 - - 3680000 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-8 Revision:

Sheet: 8 4 of 4 f. Nodal Weight and Beam Properties-Crane Girder System Node No. Nodal Weight Element No Axial Area (inches 2) 1 - 1 150 2 7400 2 150 3 7400 3 150 4 6800 4 150 5 5100 5 150 6 34700 6 150 7 34700 7 150 8 3500 8 150 9 37600 9 150 10 34000 10 150 11 5200 11 150 12 9200 12 150 13 9200 13 150 14 - 14 40 15 - 15 40 16 7700 16 40 17 7700 17 40 18 - 18 40 19 1600 19 40 20 - 20 40 21 - 21 40 22 40 23 11.8 24 11.8 25 11.8 26 11.8 27 40 28 40 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-9 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-9 CONTAINMENT S TRUCTURE S EISMIC A NALYSIS N ATURAL F REQUENCIES

, PERIODS A ND MODE C LASSIFICATION

a. Shell Mode Number Natural Frequency (Hz) Natural Period (sec.) Mode Classification 1 4.17 .2398 1st Horizontal 2 11.78 .0849 1st Vertical 3 11.84 .0845 2nd Horizontal 4 22.01 .0454 1st Rotational 5 24.17 .0414 3rd Horizontal 6 33.15 .0302 2nd Vertical 7 34.81 .0287 Higher Modes 8 40.00 .0250 Higher Modes 9 45.88 .0218 Higher Modes 10 53.28 .0188 Higher Modes 11 53.62 .0186 Higher Modes 12 59.91 .0167 Higher Modes b. Internals Mode Number Natural Frequency (Hz) Natural Period (sec.) Mode Classification 1 9.07 .1103 1st Horizontal (E-W) 2 11.90 .0840 1st Horizontal (N-S) 3 20.25 .0494 1st Torsion 4 28.37 .0353 2nd Horizontal (E-W) 5 40.09 .0249 2nd Horizontal (N-S) 6 44.88 .0223 1st Vertical 7 49.21 .0203 Higher Modes 8 73.40 .0136 Higher Modes S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-10 Revision:

Sheet: 8 1 of 3 TABLE 3.7(B)-10 CONTAINMENT S TRUCTURE S EISMIC A NALYSIS - PEAK N ODAL A CCELERATIONS

a. Peak Nodal Accelerations Due to Horizont al SSE and OBE Excitations - Containment Shell Horiz. Nodal Accelerations (g)

Safe Shutdown Earthquake Operating Basis Earthquake Mass Elevation (feet) Node Number Spectrum Method Time History Method Spectrum Method Time History Method 189.0 1 .969 1.069 .619 .692 179.5 2 .926 - .593 - 170.0 3 .879 .995 .563 .642 154.0 4 .798 .923 .513 .592 142.3 5 .741 .875 .476 .556 130.7 6 .687 - .441 - 119.0 7 .638 .769 .409 .487 105.0 8 .588 .707 .376 .450 88.5 9 .533 .637 .339 .415 72.0 10 .479 .581 .304 .373 62.5 11 .448 .545 .283 .351 43.8 12 .382 - .239 - 37.3 31 .354 .452 .222 .284 30.0 32 .326 .422 .203 .263 25.0 13 .308 - .190 - 18.0 33 .275 .370 .170 .226 12.5 14 .252 .345 .154 .209 0.0 15 .250 .294 .130 .169 -9.0 16 .250 .272 .130 .146 -18.5 34 .250 .256 .130 .131 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-10 Revision:

Sheet: 8 2 of 3 b. Peak Nodal Accelerations Due to Vertical SSE and OBE Excitations -Containment Shell Vertical. Nodal Accelerations (g)

Safe Shutdown Earthquake Operating Basis Earthquake Mass Elevation (feet) Node Number Spectrum Method Time History Method Spectrum Method Time History Method 189.0 1 .644 .565 .390 .342 179.5 2 .641 - .389 - 170.0 3 .634 .559 .385 .340 154.0 4 .614 .545 .374 .334 142.3 5 .594 .532 .362 .328 130.7 6 .570 - .348 - 119.0 7 .543 .504 .331 .311 105.0 8 .514 .491 .313 .301 88.5 9 .473 .474 .288 .287 72.0 10 .427 .452 .259 .269 62.5 11 .397 .437 .241 .257 43.8 12 .334 - .201 - 37.3 31 .308 .387 .185 .218 30.0 32 .281 .372 .168 .206 25.0 13 .262 - .156 - 18.0 33 .233 .347 .138 .190 12.5 14 .211 .336 .125 .183 0.0 15 .167 .311 .091 .167 -9.0 16 .167 .293 .083 .155 -18.5 34 .167 .273 .083 .141 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-10 Revision:

Sheet: 8 3 of 3 c. Peak Nodal Acceleration (g) - Containment Internals Earthquake Direction Node No. Elevation (ft) SSE 7% Damping Spectrum Method OBE 4% Damping Spectrum Method 29 25'-0" .249 .1246 (Vertical) 26 0'-0" .167 .0834 X 1 23 -9'-0" .167 .0830 29 25'-0" .617 .375 (N - S) 26 0'-0" .307 .181 X 2 23 -9'-0" .250 .130 29 25'-0" .726 .470 (E - W) 26 0'-0" .322 .200 Y 3 23 -9'-0" .250 .130 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-11 Revision:

Sheet: 8 1 of 3 TABLE 3.7(B)-11 CONTAINMENT S TRUCTURE S EISMIC A NALYSIS - PEAK N ODAL D ISPLACEMENTS

a. Peak Nodal Displacements Due to Horizont al SSE and OBE Excitations -Containment Shell Horizontal Nodal Displacement (inches)

Safe Shutdown Earthquake Operating Basis Earthquake Mass Elevation (feet) Node Number Spectrum Method Time History Method Spectrum Method Time History Method 189.0 1 .516 .618 .332 .393 179.5 2 .498 - .321 - 170.0 3 .479 .573 .308 .365 154.0 4 .443 .531 .285 .338 142.3 5 .415 .498 .267 .317 130.7 6 .386 - .248 - 119.0 7 .355 .427 .228 .271 105.0 8 .320 .385 .206 .244 88.5 9 .277 .334 .178 .212 72.0 10 .233 .282 .150 .179 62.5 11 .208 .252 .134 .160 43.8 12 .159 - .103 - 37.3 31 .143 .174 .092 .110 30.0 32 .125 .152 .080 .097 25.0 13 .113 - .072 - 18.0 33 .096 .118 .062 .074 12.5 14 .083 .102 .054 .065 0.0 15 .056 .069 .036 .044 -9.0 16 .038 .047 .024 .029 -18.5 34 .019 .024 .013 .015 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-11 Revision:

Sheet: 8 2 of 3 b. Peak Nodal Displacements Due to Vertic al SSE and OBE Excitations - Containment Shell Vertical Nodal Displacements (inches)

Safe Shutdown Earthquake Operating Basis Earthquake Mass Elevation (feet) Node Number Spectrum Method Time History Method Spectrum Method Time History Method 189.0 1 .0444 .0434 .0271 .0262 179.5 2 .0443 - .0270 - 170.0 3 .0439 .0430 .0268 .0259 154.0 4 .0428 .0419 .0261 .0253 142.3 5 .0416 .0408 .0254 .0247 130.7 6 .0400 - .0244 - 119.0 7 .0381 .0377 .0233 .0228 105.0 8 .0360 .0357 .0220 .0216 88.5 9 .0330 .0329 .0202 .0199 72.0 10 .0295 .0297 .0180 .0179 62.5 11 .0273 .0276 .0167 .0166 43.8 12 .0224 - .0137 - 37.3 31 .0206 .0212 .0126 .0128 30.0 32 .0185 .0192 .0113 .0115 25.0 13 .0171 - .0105 - 18.0 33 .0150 .0158 .0092 .0094 12.5 14 .0134 .0142 .0082 .0084 0.0 15 .0096 .0102 .0058 .0061 -9.0 16 .0067 .0073 .0041 .0043 -18.5 34 .0037 .0040 .0022 .0024 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS TABLE 3.7(B)-11 Revision:

Sheet: 8 3 of 3 c. Peak Nodal Displacement (inch) - Containment Internals Earthquake Direction Node No. Elevation (ft) SSE 7% Damping Spectrum Method OBE 4% Damping Spectrum Method 29 25'-0" .001181 .000591 (Vertical) 26 0'-0" .000780 .000390 X 1 23 -9'-0" .000532 .000266 29 25'-0" .04206 .02579 (N - S) 26 0'-0" .01900 .01157 X 2 23 -9'-0" .01048 .00638 29 25'-0" .08614 .05538 (E - W) 26 0'-0" .03169 .02035 X 3 23 -9'-0" .01730 .01109 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-12 Revision:

Sheet: 8 1 of 3 TABLE 3.7(B)-12 CONTAINMENT S TRUCTURE S EISMIC A NALYSIS - PEAK F ORCES a. Peak Seismic Forces - Containment Shell OBE (.13g) 4% Damping SSE(.25g) 7% Damping Beam No. Axial Force (Kips) Shear Force (Kips) Overturning Moment (10 3 Kips-in)

Axial Force (Kips) Shear Force (Kips) Overturning Moment (10 3 Kips-in) 1 455 722 100 751 1130 157 2 1363 2104 396 2247 3290 622 3 2567 3865 1281 4230 6038 2009 4 3833 5595 2262 6308 8728 3545 5 4866 6942 3428 8001 10819 5366 6 5856 8178 4772 9620 12732 7463 7 7047 9604 6631 11563 14941 10356 8 8527 11300 9162 13978 17565 14292 9 9994 12903 11992 16371 20053 18687 10 11027 13988 13783 18057 21741 21466 11 12067 15069 17312 19756 23426 26938 14 13206 16212 21112 21619 21515 32831 16 13935 16920 23744 22814 26329 36914 17 14393 17357 26378 23567 27020 41000 18 14675 17619 28294 24032 27437 43973 20 14953 17873 32744 24491 27843 50884 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-12 Revision:

Sheet: 8 2 of 3 b. Peak Forces Due to Horizontal (N-S) Excitation - Containment Internals Earthquake Elevation Beam No. Shear Forces N-S Component (Kips) Bending Moments E-W Component (Kips) OBE -26' 25 5748 2.80 -9' 27 5092 1.64 0' 30 3644 1.093 12.5' 31 8644 0.54711 SSE -26' 25 9458 4.597 -9' 27 8356 2.690 0' 30 5992 1.798 12.5' 31 5992 0.898 c. Peak Forces Due to Horizontal (E-W) Excitation - Containment Internals Earthquake Elevation Beam No. Shear Forces N-S Component (Kips) Bending Moments E-W Component (Kips) Torque 10 6 Kip-in OBE -26' 25 6685 3.352 0.4487 -9' 27 6000 2.000 0.4172 0' 30 4525 1.357 0.2252 12.5' 31 4525 0.6787 0.2252 SSE -26' 25 10454 5.215 0.699 -9' 27 9359 3.113 0.649 0' 30 7052 2.116 0.353 12.5 31 7052 1.058 0.353 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-12 Revision:

Sheet: 8 3 of 3 d. Peak Axial Forces Due to Vertical SSE and OBE Excitations By Response Spectrum Method - Containment Internals Axial Forces (Kips) Beam No. Elevation Safe Shutdown Earthquake Operating Basis Earthquake 25 -26'-0" to -9'-0" 3331 1665 27 -9'-0" to 0'-0" 2582 1290 30 0'-0" to 12'-6" 1510 755 31 12'-6" to 25'-0" 1510 755 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-13 Revision:

Sheet: 8 1 of 2 TABLE 3.7(B)-13 P RIMARY AUXILIARY B UILDING S EISMIC A NALYSIS - PEAK N ODAL a. Peak Nodal Displacements and Accelerations due to X2 (N-S) and X3 (E-W) Motions (RSS Sum)

OBE SSE Direction of Response Node No. Due to X2 Motion Due to X3 Motion SRSS Due to X2 Motion Due to X3 Motion SRSS X2 2 0.1242 0.0027 0.1242 0.1931 0.0042 0.1931 (N-S) 5 0.0862 0.0013 0.0862 0.1339 0.0020 0.1339 8 0.0508 0.0007 0.0508 0.0789 0.0011 0.0789 11 0.0147 0.0002 9.0147 0.0229 0.0003 0.0229 X3 2 0.0019 0.2572 0.2572 0.0030 0.3997 0.3997 (E-W) 5 0.0013 0.1660 0.1660 0.0020 0.2579 0.2579 8 0.0010 0.0832 0.0832 0.0015 0.1293 0.1293 11 0.0002 0.0154 0.0154 0.0003 0.0239 0.0239 X2 2 0.6682 0.0253 0.6687 1.0593 0.0438 1.0602 (N-S) 5 0.4400 0.0059 0.4400 0.6861 0.0095 0.6862 8 0.2705 0.0105 0.2707 0.4279 0.0181 0.4283 11 0.1300 0.0046 0.1301 0.2500 0.0084 0.2501 X3 2 0.0261 0.7166 0.7171 0.0449 1.1309 1.1318 (E-W) 5 0.0065 0.4406 0.4406 0.0105 0.6872 0.6873 8 0.0112 0.2423 0.2426 0.0191 0.3879 0.3884 11 0.0040 0.1300 0.1301 0.0073 0.2500 0.2501 X4 2 8.533x10-5 1.1517x10-4 1.433x10-4 1.485x10-4 2.019x10-4 2.506x10-4 (Vert.) 5 6.523x10-5 8.4598x10-5 1.068x10-4 1.076x10-4 1.433x10-4 1.742x10-4 -Torsional 8 4.756x10-5 6.3888x10-5 7.965x10-5 8.072x10-5 1.091x10-4 1.357x10-4 11 2.222x10-5 2.6195x10-5 3.435x10-5 3.851x10-5 4.711x10-5 6.085x10-5 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-13 Revision:

Sheet: 8 2 of 2 b. Peak Nodal Accelerations Due To Vertical SSE AND OBE Excitations Vertical Nodal Acceleration (g)

Elevation (Feet)

Node No. SSE OBE 108.00 1 0.39 0.21 2 0.84 0.54 3 0.33 0.18 4 0.90 0.58 81.00 5 0.60 0.38 6 0.86 0.51 7 2.75 1.48 65.75 9 0.90 0.59 10 0.24 0.14 11 0.82 0.53 53.00 12 0.70 0.45 13 1.61 1.01 14 1.51 0.85 15 0.19 0.09 16 0.80 0.51 25.00 17 1.06 0.69 18 0.69 0.43 19 0.81 0.46 7.00 20 0.70 0.45 21 0.53 0.31 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-14 Revision:

Sheet: 8 1 of 2 TABLE 3.7(B)-14 P RIMARY AUXILIARY B UILDING S EISMIC A NALYSIS - PEAK HORIZONTAL a. Peak Horizontal Responses - (RSS SUM)

Shear Forces (Kips)

Bending Moments (Kip-in)

Earthquake Direction Beam No. N-S Component E-W Component About N-S Axis About E-W Axis Torque (Kip-in) 2 1463.1 57.2 18538 474040 22189 OBE 5 5532.2 56.4 31488 2315400 815930 8 8371.0 127.0 46502 5107700 615340 X2 11 9258.9 161.5 74077 7092500 1133500 (N-S) 2 2319.5 98.3 31842 751520 37025 SSE 5 8609.8 91.4 53301 3606700 1273600 8 13015.0 208.8 72437 7936600 979160 11 14421.0 269.7 117370 11019000 1784200 2 55.5 1569.1 508830 17980 27163 OBE 5 71.5 5633.5 2382200 38119 954730 8 127.5 8079.5 5070200 59945 991020 X3 11 161.5 8594.0 6909800 86661 583090 (E-W) 2 96.0 2476.2 8023000 31097 47611 SSE 5 114.8 8764.2 3707900 63325 1492200 8 209.0 12572.0 7878400 43963 1578500 11 269.7 13409.0 10739000 136910 982020 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-14 Revision:

Sheet: 8 2 of 2 b. Maximum Axial Forces Due to Vertical SSE and OBE Excitations Axial Forces (Kips) Beam No. Safe Shutdown Earthquake Operating Basis Earthquake 1 671.46 385.70 2 2990.20 1644.40 3 2988.70 1643.70 4 4997.10 2747.50 5 5908.60 3251.30 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-15 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-15 CONTROL A ND D IESEL GENERATOR B UILDING S EISMIC A NALYSIS - N ATURAL F REQUENCIES

, PERIODS A ND MODE C LASSIFICATION Mode Number Natural Frequency (Hz) Natural Period (second) Mode Classification 1 5.6911 0.1757 Bending about X3 axis, (1 st) 2 9.0942 0.1100 Bending about X2 axis, (1 st) 3 11.6257 0.0860 Rotation about X1, (1 st) 4 17.5683 0.0569 Bending about X3, (2 nd) 5 18.4860 0.0541 Vertical, (1 st) 6 23.8661 0.0419 Bending about X2, (2 nd) 7 27.0055 0.0370 Bending about X3, (3 rd) 8 31.1205 0.0321 Rotation about X1, (2 nd) 9 33.2826 0.0300 Bending about X2, (3 rd) 10 35.6255 0.0281 Vertical, (2 nd) 11 39.6037 0.0253 Rotation about X1, (3 rd) 12 53.2650 0.0188 Higher mode 13 67.5424 0.0148 Higher mode 14 68.5735 0.0146 Higher mode 15 80.4142 0.0124 Higher mode S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-16 Revision:

Sheet: 8 1 of 2 TABLE 3.7(B)-16 CONTROL A ND D IESEL GENERATOR B UILDING S EISMIC A NALYSIS - PEAK N ODAL a. Peak Nodal Acceleration SSE 7% Damping OBE 4% Damping Time Hist Method Spectrum Method Time Hist. Method Earthquake Direction Node No. Elevation (ft) Spectrum Method (g) (g) (g) (g) 3 98.0 0.482 0.416 0.270 0.234 X1 6 78.5 0.322 0.365 0.179 0.198 VERTICAL 9 75.0 0.392 0.388 0.219 0.215 12 51.5 0.197 0.318 0.109 0.169 15 50.0 0.233 0.330 0.130 0.178 X2 3 98.0 0.868 0.942 0.556 0.568 HORIZONTAL 6 78.5 0.513 0.606 0.328 0.353 (N-S) 9 75.0 0.588 0.688 0.377 0.403 12 51.5 0.256 0.399 0.159 0.219 15 50.0 0.319 0.454 0.197 0.259 X3 3 98.0 0.731 0.695 0.468 0.430 HORIZONTAL 6 78.5 0.599 0.607 0.385 0.358 (E-W) 9 75.0 0.581 0.594 0.373 0.349 12 51.5 0.357 0.440 0.226 0.240 15 50.0 0.343 0.433 0.217 0.234 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-16 Revision:

Sheet: 8 2 of 2 a. Peak Nodal Acceleration SSE 7% Damping OBE 4% Damping Time Hist Method Spectrum Method Time Hist. Method Earthquake Direction Node No. Elevation (ft) Spectrum Method (g) (g) (g) (g) 3 98.0 0.01407 0.01373 0.00794 0.00765 X1 6 78.5 0.00948 0.00978 0.00542 0.00539 VERTICAL 9 75.0 0.01163 0.01152 0.00659 0.00640 12 51.5 0.00587 0.00619 0.00339 0.00340 15 50.0 0.00689 0.00703 0.00392 0.00389 X2 3 98.0 0.25722 0.30061 0.16554 0.17799 HORIZONTAL 6 78.5 0.14751 0.17781 0.09491 0.10448 (N-S) 9 75.0 0.17413 0.20552 0.11207 0.12135 12 51.5 0.06150 0.07703 0.03954 0.04483 15 50.0 0.07846 0.09474 0.05047 0.05561 X3 3 98.0 0.08531 0.08681 0.05480 0.05258 HORIZONTAL 6 78.5 0.07077 0.07296 0.04547 0.04384 (E-W) 9 75.0 0.06853 0.07076 0.04403 0.04249 12 51.5 0.04042 0.04290 0.02596 0.02549 15 50.0 0.03872 0.04113 0.02487 0.02443

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-17 Revision:

Sheet: 8 1 of 3 TABLE 3.7(B)-17 CONTROL A ND D IESEL GENERATOR B UILDING S EISMIC A NALYSIS - PEAK F ORCES a. Peak Forces Due to Horizontal OBE Excitation Shear Forces (Kips)

Bending Moments (Kip-in) Beam No. N-S Component E-W Component About N-S Axis About E-W Axis Torque (Kip-in) 1 1758.7 1349.2 442940 1418000 245510 2 1861.9 2439.7 1005900 1793000 285010 3 2207.1 1824.4 287120 552170 81225 4 3934.7 3132.9 739260 1863900 335790 5 4274.2 3660.3 1200900 2245800 444710 6 2201.3 1878.8 268760 571200 95508 7 2229.4 2677.1 189430 207150 704940 8 3667.9 4573.9 905730 696150 417720 9 4536.6 5017.5 420560 1053300 158910 10 5502.6 6089.0 1613100 3977000 522740 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-17 Revision:

Sheet: 8 2 of 3 b. Peak Forces Due to Horizontal SSE Excitation Shear Forces (Kips)

Bending Moments (Kip-in) Beam No. N-S Component E-W Component About N-S Axis About E-W Axis Torque (Kip-in) 1 2739.4 2100.9 700770 2205500 382510 2 2908.7 3799.0 1571500 2786600 446310 3 3441.2 2851.6 448830 861370 129190 4 6115.6 4878.1 1157100 2897700 524710 5 6651.5 5699.8 1874100 3489600 695840 6 3439.1 2938.8 420350 891580 152550 7 3475.2 4170.5 333860 323520 1095900 8 5700.6 7122.1 1415300 1082000 652050 9 7051.7 7810.8 662930 1640000 260050 10 8561.9 9481.7 2514100 6179500 817920 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-17 Revision:

Sheet: 8 3 of 3 c. Peak Axial Forces Due to V ertical SSE and OBE Excitations Axial Forces (Kips) Beam No. Safe Shutdown Earthquake Operating Basis Earthquake 1 1558.0 901.3 2 2872.7 1732.2 3 1908.2 1076.5 4 3723.1 2127.0 5 5169.2 3045.1 6 1981.3 1109.8 7 2346.1 1314.6 8 4075.4 2282.9 9 4799.0 2696.7 10 5745.1 3221.3 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-18 Revision:

Sheet: 8 1 of 3 TABLE 3.7(B)-18 F UEL STORAGE B UILDING S EISMIC A NALYSIS - N ATURAL F REQUENCIES

a. Natural Frequencies, Peri ods and Mode Classification Mode Number Natural Frequency (Hz) Mode Classification 1 10.35 1st Horizontal (N-S) 2 17.16 Mixed Mode 3 32.41 2nd Horizontal (N-S) 4 40.98 3rd Horizontal (N-S) 5 47.95 - 6 65.88 - 1 8.69 - 2 11.24 1st Horizontal (E-W) 3 18.07 Mixed Mode 4 38.64 2nd Horizontal (E-W) 5 52.10 3rd Horizontal (E-W) 6 65.62 -

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-18 Revision:

Sheet: 8 2 of 3 b. Summary of Single Degree of Freedom Systems (Vertical)

Floor Area: Roof, Filter Floor, MCC and RR Track Floor Elevation (feet) Frequency (Hz) Weight (Kips) Location (Column Lines)

Element Type 4.0 850 C to D slab/beam 5.5 1203 whole area bet. A to C, except @ 2 and 3 slab/beam/ girder 13.5 271 @ 2 and 3 84.0 4.0 563 whole area except @ 2 and 3 slab/beam/ girder 64.0 8.5 140 @ 2 and 3 beam 21.5, 20.5 9.0 178 B to C 3 to 4 slab/beam 12.0 84 B to C 3 to 4 slab/beam 10.0 430 B to C/3 to 4 A to B/2 to 3 slab/beam slab/beam 20.0 90 A to B 2 to 3 slab/beam slab/beam S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-18 Revision:

Sheet: 8 3 of 3 c. Natural Frequencies and Mode Cl assification - Crane Girder System Mode Number Natural Frequency (Hz) Mode Classification 1 10.49 1st Vertical Mode 2 18.16 2nd Vertical Mode 3 35.20 3rd Vertical Mode 4 40.34 4th Vertical Mode 5 50.34 - 6 55.63 - 7 85.93 -

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-19 Revision:

Sheet: 8 1 of 7 TABLE 3.7(B)-19 F UEL STORAGE B UILDING S EISMIC A NALYSIS - PEAK N ODAL a. Peak Nodal Acceleration Due to Horizontal Excitations - (N-S Motion)

Translational Acceleration (g)

N-S Component Earthquake Motion Mass Elevation (feet)

Node No. Spectrum Method Time History Method Rotational Acceleration About Vertical Axis (*) 84 2 .616 .641 9.10 4 .484 .541 12.00 64 5 .439 .479 8.08 SSE 6 .265 .345 9.67 46.5 7 .303 .336 5.17 8 .250 .286 5.51 34.5 9 .250 .287 5.65 25 12 .250 .253 1.73 84 2 .385 .408 5.15 4 .303 .332 6.80 64 5 .268 .285 4.37 6 .159 .200 5.13 OBE 46.5 7 .172 .192 2.78 8 .125 .155 2.88 34.5 9 .125 .154 2.88 25 12 .125 .125 8.81 * (10-5 rad/sec 2)/(386.4 in/sec

2)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-19 Revision:

Sheet: 8 2 of 7 b. Peak Nodal Acceleration Due to Horizontal Excitati ons (E-W Motion)

Translational Acceleration (g)

N-S Component Earthquake Motion Mass Elevation (feet) Node No. Spectrum Method Time History Method Rotational Acceleration About Vertical Axis (*) 84 2 .471 .494 10.84 5 .340 .382 8.85 64 6 1.333 1.148 - 46.5 8 .250 .301 7.97 SSE 34.5 9 .250 .270 6.38 25 12 .250 .253 5.79 84 2 .291 .286 6.33 5 .207 .215 5.11 64 6 .852 .710 - 46.5 8 .127 .160 4.22 OBE 34.5 9 .125 .739 3.26 25 12 .125 .128 2.90 * (10-5 rad/sec 2)/(386.4 in/sec

2)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-19 Revision:

Sheet: 8 3 of 7 c. Peak Nodal Accelerations Due to Vertical SSE and OBE Excitations Vertical Nodal Acceleration (g)

Safe Shutdown Earthquake Op erating Basis Earthquake Mass Elevation (feet) Node No. Spectrum Method Time History Method Spectrum Method Time History Method 1 .726 .758 .468 .471 84 2 .624 .846 .402 .584 3 .534 .506 .317 .296 4 .301 .297 .151 .154 5 .515 .753 .332 .466 64 6 .607 .589 .391 .397 7 .248 .287 .124 .147 46.5 8 .171 .273 .085 .140 34.5 9 .167 .261 .003 .132 25.0 10 .167 .251 .083 .127 13 .429 .461 .254 .268 21.5 14 .343 .335 .129 .186 11 .567 .606 .365 .416 20.5 12 .473 .469 .288 .271 13 .429 .461 .254 .260 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-19 Revision:

Sheet: 8 4 of 7 d. Peak Nodal Accelerations Due to Vertical SSE and OBE Excitations (Crane Girder Support System)

Vertical Nodal Acceleration (g)

Mass Elevation (feet) Node No. SSE Time History Method OBE Time History Method 2 .268 .136 3 .311 .162 4 .348 .184 5 .452 .263 6 .528 .329 64.0 7 .586 .372 8 .569 .357 9 .592 .377 10 .538 .333 11 .515 .313 12 .399 .216 13 .284 .145 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-19 Revision:

Sheet: 8 5 of 7 e. Peak Nodal Displacements Due to Horizontal SSE and OBE Excitations Horizontal Nodal Displacements (inches) N-S Component Earthquake Motion Elevation (feet) Mass Node No. Spectrum Method Time History Method 84 2 .0563 .0615 4 .0440 .0486 64 5 .0372 .0413 6 .0215 .0242 SSE 46.5 7 .0195 .0221 8 .0095 .0108 34.5 9 .0087 .0100 25 12 .0012 .0014 84 2 .0351 .0382 4 .0275 .0300 64 5 .0233 .0255 6 .0135 .0149 OBE 46.5 7 .0122 .0135 8 .0059 .0066 34.5 9 .0054 .0061 25 12 .0008 .0009 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-19 Revision:

Sheet: 8 6 of 7 f. Peak Nodal Displacements Due to Horizontal SSE and OBE Excitations Horizontal Nodal Displacements (inches) E-W Component Earthquake Motion Elevation (feet) Mass Node No. Spectrum Method Time History Method SSE 84 2 .0413 .0462 5 .0284 .0322 64 6 .1628 .1517 46.5 8 .0145 .0169 34.5 9 .0064 .0077 25 12 .0009 .0012 OBE 84 2 .0258 .0271 5 .0178 .0187 64 6 .1045 .0960 46.5 8 .0091 .0099 34.5 9 .0040 .0045 25 12 .0006 .0007 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-19 Revision:

Sheet: 8 7 of 7 g. Peak Nodal Displacements Due to Vertical SSE and OBE Excitations Vertical Nodal Displacements (inches)

Safe Shutdown Earthquake Operating Basis Earthquake Mass Elevation (feet) Node No. Spectrum Method Time History Method Spectrum Method Time History Method 84 4 .0031 .0047 .0018 .0026 64 7 .0025 .0037 .0014 .0021 46.5 8 .0015 .0023 .0008 .0013 34.5 9 .0007 .0012 .0004 .0007 25 10 .0001 .00019 .00006 .0001 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-20 Revision:

Sheet: 8 1 of 2 TABLE 3.7(B)-20 F UEL STORAGE B UILDING S EISMIC A NALYSIS - MAXIMUM F ORCES a. Maximum Forces Due to Horizontal Excitations Beam No. Shear Forces (Kips)

Bending Moments (Kip-in)

Torque (10 5 Kip-in) Earthquake Motion N-S E-W N-S E-W About N-S Axis About E-W Axis N-S E-W SSE 1 1 1140 3126 750320 326920 2.438 4.377 2 6 1977 3126 750320 419470 1.958 - 3 2 2683 4685 1730400 889300 4.578 8.095 4 7 2295 4685 1730400 901070 1.998 - 5 3 2904 5144 2463600 1307100 4.706 7.386 6 7 2546 5144 2463600 1266000 2.073 - 7 4 2972 5297 3070500 1645600 4.752 7.516 8 7 2637 5297 3070500 1565000 2.106 - 9 5 5689 5334 3357200 3513800 6.851 5.278 OBE 1 1 712 1949 467650 204540 1.498 2.709 2 6 1236 1949 467650 262560 1.225 - 3 2 1680 2926 1080600 556990 2.866 5.03 4 7 1437 2926 1080600 564240 1.249 - 5 3 1819 3213 154220 818770 2.945 4.573 6 7 1594 3213 154220 793030 1.294 - 7 4 1862 3307 1918300 1030800 2.973 4.648 8 7 1649 3307 1918300 980290 1.314 - 9 5 3524 3327 2097500 2201100 4.286 3.218 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.7(B)-20 Revision:

Sheet: 8 2 of 2 b. Maximum Axial Forces Due to Vertical SSE and OBE Excitations Axial Forces (Kips) Beam No. Safe Shutdown Earthquake Op erating Basis Earthquake 1 1276 755 2 1985 1106 3 2245 1225 4 2345 1271 5 2367 1282 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-21 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-21 F REQUENCY I NTERVAL F OR C ALCULATING FLOOR RESPONSE S PECTRA FREQUENCY RANGE, HZ.

INCREMENT 0.5 - 1.6 0.1 1.6 - 2.8 0.2 2.8 - 4.0 0.3 4.0 - 9.0 0.5 9.0 - 16.0 1.0 16.0 - 22.0 2.0 22.0 - 37.0 3.0 37.0 - 45.0 4.0 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-22 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-22 N ON-C ATEGORY I S TRUCTURES D ESIGNED AGAINST COLLAPSE O NTO A DJACENT C ATEGORY I S TRUCTURES D UE T O SSE L OADS Non-Category I Structure Affected Category I Structure Turbine Building (1) Condensate Storage Tank (E.W.) Containment and Others (N.S.) Nonessential Switchgear Building (2) Control & Diesel Generator Building Tank Farm Area (Steel Framing Portion which Includes Steel Framing, Concrete Roofing and Metal Siding over Refueling Water Storage Tank)

(3) Primary Auxiliary Building, Waste Processing Building and Tank Farm Area Tunnels Steam Generator Blowdown Recovery Building (4) Primary Auxiliary Building, Waste Processing Building and Tank Farm Circulating Water Pumphouse Steel Framing Portion (5) Service Water Pumphouse Waste Processing Building Steel Framing Portion and

Reinforced Concrete Porti on, Except the Area Between Columns 1 to 2 and A to D Between Elevations 53'-0" and 86'-0" (6) Primary Auxiliary Building Tank Farm Area and Piping Tunnels NOTES: (1) The entire Turbine Building is designed against failure in the north-south direction. The south end is designed against failure in the east-west direction; and east-west failure in the north end will not affect any seismic Category I structures.

(2) The Nonessential Switchgear Building is designed mechanistically to fall away from the Control and Diesel Generator Building under the action of a collapsing Administration and Service Building. Thus, no significant load is applied to the Control and Diesel Generator Building by either the falling Administration and Service Building or the falling Nonessential Switchgear Building.

(3) The steel framing portion is designed and constructed such that the operating basis earthquake (OBE) and safe shutdown earthquake (SSE) would not cause the steel framing portion to collapse upon any safety-related structures, systems or components within or surrounding the Tank Farm area.

(4) The Steam Generator Blowdown Recovery Building is designed not to collapse.

(5) The collapse of the Circulating Water Pumphouse on the Service Water Pumphouse was evaluated to prove that the collapse will not impair the Service Water Pumphouse or system.

(6) The Waste Processing Building is designed not to collapse.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-23 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-23 C RITICAL D AMPING V ALUES Item, Equipment or Component Damping Percent Critica OBE SSE Piping Systems 1 2 Valves, Compact Pumps, Compresso rs, Diesel Generators, Pipe Mounted Instrumentation 1 2 Heat Exchangers, Tanks & Vessels, Control Cabinets, Deep Well Pumps, Fans, Electrical Switchgear, Filters, Dampers, Motors 2 3 Electrical Conduits with Bolted Connections 4 7 Electrical Conduit with WeldedConnections 2 4 Equipment Supported bySeismic Dampers 8 8 Cable Trays (Refer to Subsection 3.7(B).1.3) NOTE: Higher damping values may be used provided that adequate justification is available

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(B)-24 Revision:

Sheet: 8 1 of 1 TABLE 3.7(B)-24 S EISMIC INSTRUMENTATION L OCATION (S) Item Seismic Instrumentation Location Location Figure 1 Triaxial Time History Accelerograph, XT-6700 Free-field position in control room intake, on

bedrock 3.7(B)-32 2 Triaxial Time History Accelerograph, XT-6701 Containment building foundation between Columns 16 and 17 1.2-2 3 Triaxial Time History Accelerograph, XT-6710 Containment Building between Columns 16 and 17 on the concrete operating floor 1.2-4 4 Triaxial Time History Accelerograph, XR-5707 Primary Auxiliary Building, elevation 53'-0" 1.2-9 5 Triaxial Time History Accelerograph, XR-6708 Service water pumphouse electrical room, west wall, north of column 12, at elevation 22'-0" 1.2-46 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.7(N)-1 Revision:

Sheet: 8 1 of 1 TABLE 3.7(N)-1 D AMPING V ALUES U SED B Y WESTINGHOUSE F OR S EISMIC S YSTEMS A NALYSIS Damping (Percent of Critical)

Item Upset Conditions (OBE) Faulted Condition (SSE, DBA) Primary coolant loop system components and large piping a 2* 4* Small piping 1* 2* Welded steel structures 2 4 Bolted and/or riveted steel structures 4 7 aApplicable to 12 inch or larger diameter piping. *The damping values of ASME B&PV Code, Code Case N-411, may be used for pipe stress verification and for pipe support optimization in place of the valu es given in Regulatory Guide 1.61.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-1 Revision:

Sheet: 8 1 of 2 TABLE 3.8-1 CONTAINMENT L OAD C OMBINATIONS A ND L OAD F ACTORS (5) LOADING (6) R r (DBA Local Effects)

Category Load Combination Number Dead Load Live Load Test Pressure Accident Pressure Test Temperature Normal Temperature DBA Temperature Operating Basis Earthquake Safe Shutdown Earthquake Wind Load Tornado Normal Pipe Reaction DBA Thermal Pipe Reaction Reaction of Ruptured High Energy Pipe Jet Impingement Loads Impact of Ruptured High Energy Pipe Pressure Variations Design Basis Flood Design Conditions Loading Notation D L P t P T E R R R R R P F a T t o T a (1) E o ss W W t o a rr rj rm v (2) (3) Test 1 1.0 1.0 1.0 - 1.0 - - - - - - - - - - - - - Normal 2 1.0 1.0 - - - 1.0 - - - - - 1.0 - - - - 1.0 - Service Load Severe Environmental 3 1.0 1.0 - - - 1.0 - 1.0 - - - 1.0 - - - - 1.0 - Severe Environmental 4 1.0 1.3 - - - 1.0 - 1.5 - - - 1.0 - - - - 1.0 - 5a 1.0 1.0 - - - 1.0 - - - - 1.0 (7) 1.0 - - - - 1.0 1.0 Extreme Environmental 5b 1.0 1.0 - - - 1.0 - - 1.0 - 1.0 (7,8) 1.0 - - - - 1.0 - 6a 1.0 1.0 - 1.5 (4) - - 1.0 (4) - - - - - 1.0 - - - - - Abnormal 6b 1.0 1.0 - 1.0 (4) - - 1.0 (4) - - - - - 1.25 - - - - - Abnormal/Severe Environmental 7 1.0 1.0 - 1.25 (4) - - 1.0 (4) 1.25 - - - - 1.0 1.0 1.0 1.0 - - Factored Load Abnormal/Extreme Environmental 8 1.0 1.0 - 1.0 (4) - - 1.0 (4) - 1.0 - - 1.0 1.0 1.0 1.0 - -

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-1 Revision:

Sheet: 8 2 of 2 (1) Includes effect of normal operating therma l loads and accident loads. For all abnormal load conditions, structure should be c hecked to assure that accident pressure load without therma l load can be resisted by the structure within the specified allowable stresses for this condition.

(2) Negative pressure variations inside the structure shall not be considered simultaneously with outside negative pressure due to tornado loadings.

(3) For this load case, the design basis flood elevation shall be the max. ground water elevation, i.e., El. +20'-0".

(4) Load cases examined for maximum pressure and its coincident liner temperature and maximum liner temperature with its coinciden t pressure.

(5) All load factors shall be taken as 1.0 for the design of steel liner.

(6) See Subsection 3.8.1.3 for discussion of loadings.

(7) W t includes missile effects only.

(8) For this load case, loadings from E ss or W t included individually.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.8-2 Revision:

Sheet: 8 1 of 2 TABLE 3.8-2 ALLOWABLE S TRESSES A ND S TRAINS I N THE CONTAINMENT S TRUCTURE SERVICE LOADS FACTORED LOADS Mechanical Loads Thermal or Test Loads Added All Loads fc' = 4000psi fc' = 3000psi Concrete Compressive Stresses Membrane Memb. + Bending

Membrane Memb. + Bending Shear Stresses Radial

Tangential

Punching 0.9 ksi 1.35 ksi

1.2 ksi 1.8 ksi CC-3431.3 & Code CASE N-287 (2)

CC-3431.3 CC3431.3 & Code Case

N-219 (3) 1.35 ksi 1.8 ksi

1.8 ksi 2.4 ksi CC-3431.3 & Code CASE N-287 (2) CC-3431.3

CC3431.3 & Code Case N-219 (3) 2.25 ksi 2.55 ksi

3.0 ksi 3.4 ksi CC-3431.4.1 & Code Case N-287 (2) CC-3421.5.1

CC-3421.6 & Code Case N-219 (3)

Reinforcing (fy = 60 ksi)

Tensile stress 30.0 ksi 45.0 ksi 40.0 ksi (4) 54.0 ksi (5)

Liner (6) Tensile Strain Membrane Memb. + Bending Compressive Strain Membrane Memb. + Bending

.002 in/in

.004 in/in

.002 in/in

.004 in/in

.002 in/in

.004 in/in

.002 in/in

.004 in/in

.003 in/in

.010 in/in

.005 in/in

.014 in/in Liner Anchorage System Force/displacement limits are based on tests which are desc ribed in Appendix 3G and are in accordance with Table CC-3730-1

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.8-2 Revision:

Sheet: 8 2 of 2 NOTES: (1) The allowable shear stress, c, to be resisted by the concrete does not exceed 40 psi and 60 psi for load combinations 7 and 8, respectively, in Table 3.8-1. (2) Radial shear reinforcement was designed per Code Case N-287. (3) Peripheral shear reinforcement was designed per Code Case N-219. (4) The allowable stresses for reinforcing steel are per Subarticle CC-3432.1, 1979 Summer Addendum. (5) Stresses and strains in reinforcing steel are per Subarticle 3422.1, 1977 Winter Addendum. (6) The liner allowables are per Table CC-3720-1, 1976 Winter Addendum.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.8-3 Revision:

Sheet: 8 1 of 1 TABLE 3.8-3 P ROPERTIES O F M ATERIALS F OR DESIGN O F CONTAINMENT Concrete - 3000 psi 28-day Compressive Strength = 3000 psi Modulus of Elasticity = 3.12x10 6 psi Coefficient of Thermal Expansion = 6.5x10

-6 in./in./ F Poissons' Ratio = 0.15 Concrete - 4000 psi 28-day Compressive Strength = 4000 psi Modulus of Elasticity = 3.61x10 6 psi Coefficient of Thermal Expansion = 6.5x10

-6 in./in./ F Poissons' Ratio = 0.15 Reinforcing Steel Design Yield Strength = 60,000 psi Modulus of Elasticity = 29x10 6 psi Coefficient of Thermal Expansion = 6.5x10

-6 in./in./ F Containment Liner Design Yield Strength = 32,000 psi Modulus of Elasticity = 29x10 6 psi Coefficient of Thermal Expansion = 6.5x10

-6 in./in./ F Poissons' Ratio = 0.3 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.8-4 Revision:

Sheet: 8 1 of 1 TABLE 3.8-4 CONTAINMENT L OAD C ASES , B OUNDARY C ONDITIONS , C RACK P ATTERNS Description of Load Case Cracking Pattern Bound Cond. Liner Incl. in Stress Comp. Liner Incl. in Force & Moment Comp. Ref. Temp. PA = 52 psi Wall & Dome-Cracked; Mat -

Uncracked Base Z-Fixed

Base R-Free No No -- Dead Load Wall & Dome-Cracked; Mat -

Uncracked Base Z-Fixed

Base R-Free No No -- To (Operating Temp. Gradient)

(120F, inside to

(-) 10F outside) Dome-Wall Mat

Inner Half -

Uncracked

Outer Half - Cracked Base Z-Fixed Base R-Free Yes No 70 F Hot Liner (Accident Temp.

Spike) T = 151.1 F Wall & Dome -

Cracked Mat -

Uncracked Base Z-Fixed

Base R-Free Yes No 70 F OBE Shell Model -

Uncracked Bot. Of Wall -

Fixed -- -- -- SSE Shell Model -

Uncracked Bot. Of Wall -

Fixed -- -- --

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.8-5 Revision:

Sheet: 8 1 of 1 TABLE 3.8-5 B OUNDARY C ONDITIONS F OR EQUIPMENT H ATCH A NALYSIS LOAD CASE X-Z PLANE Y-Z PL ANE X-Z PLANE Y-Z PLANE Dead Load Sym. Sym. v= 2=0 u= 1=0 Pressure Sym. Sym. v= 2=0 u= 1=0 Vert. (OBE) Sym. Sym. v= 2=0 u= 1=0 Vert. (SSE) Sym. Sym. v= 2=0 u= 1=0 N-S (OBE) Sym. Anti-Sym.

v= 2=0 v=w= 2=0 N-S (OBE) Sym. Anti-Sym.

V= 2=0 v=w= 2=0 E=W (OBE) Anti-Sym. Sym. u=w= 1=0 v= 1=0 E-W (SSE) Anti-Sym. Sym. u=w= 1=0 U= 1=0 NOTE: Base is fixed i.e., u=v=w= 1= 2=0 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-6 Revision:

Sheet: 8 1 of 2 TABLE 3.8-6 EQUIPMENT H ATCH A ND PERSONNEL L OCKS - L OAD C OMBINATIONS A ND L OAD F ACTORS LOADING Category Load Combination No. Dead Load Live Load Test Pressure Accident Pressure Test Temperature Operational Temperature Accident Temperature Operating Basis Earthquake Safe Shutdown Earthquake Pressure Variations Loading Notation D L P t P T T T a t o a E E' P e Test 1 1.0 1.0 1.0 --- 1.0 --- --- --- --- --- Normal 2 1.0 1.0 --- --- --- 1.0 --- --- --- 1.0 Upset (Severe Environmental) 3 1.0 1.0 --- --- --- 1.0 --- 1.0 --- --- Faulted (Abnormal/

Severe Environmental) 4 1.0 1.0 --- 1.0 --- --- 1.0 1.0 --- --- Faulted(Abnormal/

Extreme Environmental) 5 1.0 1.0 --- 1.0 --- --- 1.0 --- 1.0 ---

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-6 Revision:

Sheet: 8 2 of 2 NOTES: 1. Number of load cycles for test condition shall be as per 10 CFR 50 Appendix J.

2. The fatigue analysis is based on the consideration of 20 occurrences of Opera ting Basis Earthquake (OBE) each having 20 cycles of maximum response per seismic event. 3. The design of equipment hatch and personnel air locks are based on Tables 3.8-6 and 3.8-10 which are in agreement with the load combinations and stress limits defined in SRP 3.8.2, Revision 0, 11/14/75. It has been shown that the design of equipment hatch and personn el air locks also meets the current (Rev. 1, 7/81) SRP requirements.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.8-7 Revision:

Sheet: 8 1 of 1 TABLE 3.8-7 L OAD C OMBINATIONS F OR DESIGN L OAD S ET A ND F OR A CTUAL A PPLIED P IPE L OADS F OR MODERATE E NERGY PIPING P ENETRATIONS Design Condition Load Combinations For Actual Pipe Loads Applied to Penetration Design/Norma P + D + T Upset P + D + T + OBE + SAD(OBE) + TH Emergency/Faulted P + D + T + SSE + SAD(SSE) + TH + PAD + TAD P Maximum Containment Pressure D Dead Weight T Thermal Load OBE Operating Basis Earthquake SAD(OBE) Seismic Anchor Displacement due to OBE SSE Safe Shutdown Earthquake SAD(SSE) Seismic Anchor Displacement due to SSE TH Thrust PAD + TAD -Pressure Plus Thermal Anchor Displacement

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT A ND SYSTEMS T ABLE 3.8-8 Revision:

Sheet: 8 1 of 1 TABLE 3.8-8 E LECTRICAL P ENETRATION A SSEMBLY L OAD C OMBINATIONS A ND L OAD F ACTORS CATEGORY Load Combination No. Dead Load Test Pressure Accident Pressure Test Temperature Operational Temperature Accident Temperature Operating Basis Earthquake Safe Shutdown Earthquake External Pressure LOADING NOTATION D P t P a T t T o T a E E' P e TEST a 1.0 1.0 -- 1.0 -- -- -- -- -- NORMAL b 1.0 -- -- -- 1.0 -- -- -- 1.0 UPSET (Severe Environmental) c 1.0 -- -- -- 1.0 -- 1.0 -- -- FAULTED (Abnormal/Severe Environmental) d 1.0 -- 1.0 -- -- 1.0 1.0 -- -- FAULTED (Abnormal/Extreme Environmental) e 1.0 -- 1.0 -- -- 1.0 -- 1.0 -- NOTES: 1. Number of load cycles for test condition shall be as per 10 CFR 50 Appendix J. 2. The fatigue analysis is based on the consideration of 20 occurrences of operating basis earthquake (OBE), each having 20 cycles of maximum response per seismic event.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-9 Revision:

Sheet: 8 1 of 1 TABLE 3.8-9 F UEL T RANSFER T UBE A SSEMBLY L OAD C OMBINATIONS A ND L OAD F ACTORS CATEGORY Load Combination No. Dead Load Test Pressure Operating Pressure Accident Pressure Test Temperature Operational/Refue ling Temperature Accident Temperature Operating Basis Earthquake Safe Shutdown Earthquake Live Load Bellow Loads Hydrostatic Pressure LOADING NOTATION D P t P o P a T T T E E t o a o s L R H TEST 1 1.0 1.0 -- -- 1.0 -- -- -- -- -- 1.0 -- NORMAL 2 1.0 -- 1.0 -- -- 1.0 -- -- -- -- -- -- NORMAL (Refueling) 3 1.0 -- -- -- -- 1.0 -- -- -- 1.0 -- 1.0 UPSET (Severe Environmental) 4 1.0 -- -- -- -- 1.0 -- 1.0 -- -- 1.0 -- FAULTED (Abnormal/Extreme Environmental) 5 1.0 -- -- 1.0 -- -- 1.0 -- 1.0 -- 1.0 -- NOTES:

1. The number of load cycles for test condition shall be as per 10 CFR 50 Appendix J.
2. The fatigue analysis is based on the consideration of 20 occurrences of the Op erating Basis Earthquake, each occurrence having 20 cycles of maximum response.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-10 Revision:

Sheet: 8 1 of 2 TABLE 3.8-10 S TRESS L IMITS F OR EQUIPMENT H ATCH A ND PERSONNEL L OCKS Primary Stresses Load Combination No General Membrane P m Local Membrane P L Bending + Loca Membrane P B + P L Primary &

Secondary Stresses Peak Stresses Buckling TEST 1 0.9 S y 1.25 S y 1.25 S y 3 S m Consider for Fatigue Analysis 125% of Allow. Given by NE-3133 CONSTRUCTION 2 S m 1.5 S m 1.5 S m 3 S m Consider for Fatigue Analysis Allow. Given By NE-3133 NORMAL 3 S m 1.5 S m 1.5 S m 3 S m Consider for Fatigue Analysis Allow. Given By NE-3133 UPSET (SEVERE

ENVIRONMENTAL) 4 S m 1.5 S m 1.5 S m 3 S m Consider for Fatigue Analysis Allow. Given By NE-3133 FAULTED (ABNORMAL/

SEVERE ENVIRONMENTAL) 5 S m 1.5 S m 1.5 S m N/A N/A Allow. Given By NE-3133 6a* S m 1.5 S m 1.5 S m N/A N/A Allow. Given By NE-3133 FAULTED (ABNORMAL/

EXTREME ENVIRONMENTAL 6b** The Greater of 1.2 S m or S y The Greater

of 1.8 S m or 1.5 S y The Greater

of 1.8 S m or 1.5 S y N/A N/A 120% Allow. Given NE-3133 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-10 Revision:

Sheet: 8 2 of 2 Primary Stresses Load Combination No General Membrane P m Local Membrane P L Bending + Loca Membrane P B + P L Primary &

Secondary Stresses Peak Stresses Buckling NOTES: (1) Thermal stresses need not be considered in computing P m , P L and P B. (2) Thermal effects are considered in: (a) Specifying stress intensity limits as a function of temperature.

(b) Analyzing effects of cyclic operation.

(3) * - Not Integral and Continuous

    • - Integral and Continuous (4) S y = yield stress S m = allowable primary membrane stress

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.8-11 Revision:

Sheet: 8 1 of 2 TABLE 3.8-11 PROPERTIES O F MATERIALS FOR PERSONNEL AIR LOCKS A ND EQUIPMENT HATCH 1. Plates a. Personnel Air Lock

- Pressure retaining areas and welded pipe SA-516, Gr. 70 Properties @ 271 F: E = 27.4x10 6 psi S y = 33,961 psi S m = 19,300 psi Sbr = 33,961 psi S s = 15,440 psi SA-283-C Properties @ 120 F: S y = 30,000 psi S m = 12,600 psi Sbr = S y = 30,000 psi S s = 0.8 S m = 10,800 psi

b. Personnel Air Lock

- Tapered wedges for door locking SA-240, Type 304 Properties @ 271 F: S y = 30,000 psi (110 F) S y = 23,225 psi Sbr = S y = 23,225 psi S m = 18,700 psi

c. Equipment Hatch

- Pressure retaining areas and welded pipe SA-516, Gr. 60 Properties @ 271 F: SA-516, Gr. 60 Properties @ 120 F: E = 27.49x10 6 E = 27.82x10 6 psi S y = 28,560 psi S y = 31,440 psi S m = 16,500 psi S m = 16,500 psi

-6 in/in/ F -6 in/in/ F 2. Bolts SA-193, Gr. B7 Properties @ 271 F: S y = 105,000 psi (110 F) S y = 94,100 psi S m = 27,500 psi S s = 22,000 psi Sbr= 94,100 psi

3. Pins C1045 Properties @ 120 F: S y = 59,000 ps i S m = 0.6S y = 35,400 psi S s = 0.4S y = 23,600 psi Sbr = 0.9S y = 53,100 psi (18-8) Type 304 (PAL* in Equipment Hatch) Properties @ 271 F: S y = 30,000 psi (110 F) S y = 22,500 psi S m = 18,300 psi Sbr = S y = 22,500 psi

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.8-11 Revision:

Sheet: 8 2 of 2 4. Pipes SA-333, Gr. 6 Properties @ 271 F: S y = 31,261 psi S m = 15,000 psi S s = 12,000 psi Sbr = S y = 31,261 psi

5. Seals Silicone or approved equal
6. Welding Filler Materials Welding Process ASME Filler Metal Specification SMAW SFA-5.1 Class E7018 SFA-5.1 Class E6010 SFA-5.4 Class E309 SAW SFA-5.17 Electrode Class

-ELX-ELXX -EMXX-EMXXX -EHXX Flux Class

-F6X-xxx -F7X-xxx GMAW-GTAW SFA-5.18 Electrode Class

-E705-X SFA-5.9 Electrode Class

-FR-309 *PAL - Personnel Air Lock

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.8-12 Revision:

Sheet: 8 1 of 1 TABLE 3.8-12 ALLOWABLE STRESS INTENSITIES FOR HIGH ENERGY PIPING PENETRATIONS Condition Loading Allowable Stress Intensity(1) Design (including normal, upset and emergency conditions)(2) Pipe operating pressure + mechanical loads (including seismic)

P m Sh Pipe operating pressure + mechanical loads (including seismic)

P 1 + P b 1.5 Sh Pipe operating pressure and temperature + mechanical loads (including seismic)

P 1 + P b + Q 3 Sh Faulted Pipe operating pressure + mechanical loads (including seismic)

P m min. of 2.4 Sh or 0.7 Su Pipe operating pressure + mechanical loads(including seismic)

P 1 + P b min. of 3.6 Sh or 1.05 Su

(1) The symbols P m , P 1 , P b, Sh and Su are defined in the ASME B&PV Code,Section III, Division 1, Subsection NA.

(2) Concrete temperatures are checked for compliance with the limits of the ASME B&PV Code,Section III, Division 2, Subsection CC.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.8-13 Revision:

Sheet: 8 1 of 2 TABLE 3.8-13 M ATERIALS F OR PIPING P ENETRATIONS

, E LECTRICAL P ENETRATIONS

, F UEL T RANSFER T UBE A SSEMBLY A ND VENTILATION P ENETRATION A SSEMBLIES 1. HIGH ENERGY PIPING PENETRATIONS

a. Process Pipe and Flued Head SA105 or SA182-304 or SA182-316
b. Sleeve SA333, Gr.1 2. MODERATE ENERGY PIPING PENETRATIONS
a. Process Pipe

SA312-304 or SA106, Gr.B or SA376-316

b. End Plate SA516, Gr.60
c. Sleeve SA333, Gr.1
3. ELECTRICAL PENETRATIONS
a. Installation Weld SA333, Gr.6
b. Monitoring Plate (Bulkhead) and Clamps

SA240-304

c. Module

SA479-304

d. Studs and Bolts SA193-B7 e. Other Materials An epoxy sealant and "O" ring seal are also used in the electrical penetrations.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.8-13 Revision:

Sheet: 8 2 of 2 4. FUEL TRANSFER TUBE ASSEMBLY

a. Bolting Flange SA182-304
b. Bolts SA164, Gr.400, Cond. A, H.F.
c. Tube SA240-304
d. Quick Closure Hatch, Mounting Ring and Slip-on Flange SA182-304 e. Other Materials Two self-energizing elastomer quad rings are also used in the fuel transfer tube assembly. 5. VENTILATION PENETRATION ASSEMBLIES
a. Sleeve SA333, Gr. 1
b. Flange

SA105

c. CAP Penetration Blind Flanges SA516, Gr. 70
d. Studs and Nuts SA 193, Grade B7 SA 194, Grade 7, 2H, 3, or 4 e. Other Materials A resilient seal material (EPDM) is used for the blind flange o-rings.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-14 Revision:

Sheet: 8 1 of 2 TABLE 3.8-14 INTERIOR CONTAINMENT S TRUCTURES B ASIC L OAD C OMBINATIONS A ND L OAD F ACTORS Material LOADING (1) Dead Load and Hydrostatic Load Live Load Accident Pressure Operational Temperature Accident Temperature Operating Basis Earthquake Safe Shutdown Earthquake Operational Piping Loads Accident Piping Loads Jet Force Reaction Jet Impingement Loads Missile Impact Loads Internal Missile Loads Stress Limit or Design Criteria Design Conditions Loading Notations Load Case Number D L P a T T E E R R R R o a o ss o a rr rj R rm M 1S 1.0 1.0 - - - - - - - - - - - 2S 1.0 1.0 - - - 1.0 - - - - - - - 3S 0.67 0.67 - 0.67 - - - 0.67 - - - - - Structural Steel 4S 0.67 0.67 - 0.67 - 0.67 - 0.67 - - - - - F s Per AISC 1C 1.4 1.7 - - - - - - - - - - - 2C 1.4 1.7 - - - 1.9 - - - - - - - 3C - - - - - - - - - - - - - Normal Load

Concrete 4C 1.05 1.28 - 1.28 - 1.43 - 1.28 - - - - - ACI 318-71 5S 0.63 0.63 - 0.63 - - 0.63 0.63 - - - - - 6S 0.63 0.63 0.63 - 0.63 - - - 0.63 - - - - 7S 0.63 0.63 0.63 - 0.63 0.63 - - 0.63 0.63 0.63 0.63 0.63 Elastic 8S 0.59 0.59 0.59 - 0.59 - 0.59 - 0.59 0.59 0.59 0.59 0.59 F s Per AISC 5S 1.1 1.1 - 1.1 - - 1.1 1.1 - - - - - 6S 1.1 1.1 1.7 - 1.1 - - - 1.1 - - - - AISC, Part II 7S 1.1 1.1 1.4 - 1.1 1.4 - - 1.1 1.1 1.1 1.1 1.1 (3) (4)

Structural Steel

(3) (4) Plastic 8S 1.1 1.1 1.1 - 1.1 - 1.1 - 1.1 1.1 1.1 1.1 1.1 5C 1.0 1.0 - 1.0 - - 1.0 1.0 - - - - - 6C 1.0 1.0 1.5 - 1.0 - - - 1.0 - - - - 7C 1.0 1.0 1.25 - 1.0 1.25 - - 1.0 1.0 1.0 1.0 1.1 Unusual Load Concrete

(4) 8C 1.0 1.0 1.0 - 1.0 - 1.0 - 1.0 1.0 1.0 1.0 1.0 ACI 318-71 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-14 Revision:

Sheet: 8 2 of 2 (F s = Allowable Stress)

(1) See Subsection 3.8.3.3 for discussion of loadings.

(2) In above load combinations, the peak values of P a , T a , R a , R rj , R rr , R rm and M shall be combined (when they act concurrently) unless time history analysis is performed to justify otherwise.

(3) For these load combinations either elastic or plastic design may be used.

(4) Load combinations 7S, 8S, 7C and 8C are also checked without R rr , R rj , R rm.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-15 Revision:

Sheet: 8 1 of 3 TABLE 3.8-15 D ESCRIPTIONS OF S EISMIC C ATEGORY I S TRUCTURES Foundation Building Structure Length Width Thickness Comment s Length Width Height Comments Figures Containment Enclosure Building - 10'-3" 10' Ring with Inside Diameter = 153'-0" & Outside Diameter =

173'-6" - 162'-6" 230' A 1.2-2 through1.2-6 Containment Equipment Hatch Missile

Shield 1.2-4, 1.2-5 Containment Structure - 153' diam 10' Also includes pit for reactor - 140' I.D. 219' A 1.2-2 through 1.2-6 Containment Internal Structures See Containment Structure See Subsection 3.8.3 1.2-2 through 1.2-6 Containment Enclosure Ventilation Area 116'-0" 55'-9" Maximum 2'-6" Overall Dimensions 116'-0" 55'-9" Maximum 31'-6" Overall dimensions (Pipe tunnel below not

included) 1.2-2 through1.2-4 Control and Diesel Generator Building 138'-0" A 95'-0" A 90'-0" 92'-0" 4', 3'-6"

& 5'-5" Stepped fdn. CB Stepped fdn. DGB 138'-0" 95'-0" 90'-0" 90'-0" 119'-3" A 95"-0" A 1.2-31 through 1.2-23 CB1.2-34 through1.2-36

DGB Control Room Makeup Air Intake Structur e 20'-4" 20'-4" 3'-0" - 18'-4" 18'-4" 10'-9" 3.8-31 Emergency Feedwater Pump Building including Electric Cable Tunnels and Penetration Areas (Control Building to Containment) 86' 44'-61/4" maximum 4' Overall Dimensions 84' 41'-31/4" 73' 1.2-2 through Enclosure for Condensate Storage Tank 29'-0" diam.

5'-0" 21'-3 I.D.

Shield wall

around tank 3.8-32 Fuel Storage Building 100' A 98'-6" 2' to 4'-43/4" Stepped fdn.

98' 97'-6" 110' A 1.2-15 through 1.2-21 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-15 Revision:

Sheet: 8 2 of 3 Foundation Building Structure Length Width Thickness Comments Length Width Height Comments Figures Main Steam and Feedwate r Pipe Chase (East) including East Penetration Area 133'-7" 30'-3" maximum 6'-0" 127'-7" 22'-3" maximum 61'-6" 1.2-3, 1.2-4 Main Steam and Feedwater Pipe Chase (West) including Mechanical Penetration Area and Personnel Hatch

Area 127'-9"

45'-9" 34'-81/2"

37'-0" maximum 6'-0" 13'-0" Chase

Hatch Area (Fill Conc) 123'-9"

45'-9" 20'-0"

37'-0" 99'-0" maximum 29'-3" Chase and Mechanical Penetration

Area Hatch Area (irregular

shape) 1.2-2 through 1.2-5 Piping Tunnels 192'15' approx.

total 2'-3" min. Thickness varies 192' approx.

total 14' 10'-11" min. Height varies 3.8-33 Pre-Action Valve

Building 47'-10" 35'-0" 2'-2" to 4'-0" Stepped fdn. overall dimensions 34'-10" 34'-6" 27'-4" Overall dimensions 1.2-51 Primary Auxiliary

Building including Residual Heat Removal Equipment Vault 147' A 62'-6" 81' A 48'-6" 4' 4' Stepped fdn.

Actually part of PAB foundation 145' 57'-0" 79' 43'-0" 101' 86'-6" PAB RHR Equip.

Vault 1.2-9 through 1.2-12 1.2-13, 1.2-14 Safety-Related

Electrical Duct Banks and Manholes 22'-6" to 24'-0" 19'-0" to 23'-6" 2'-6" Manholes-overall dimensions-several sizes used for

Duct banks 18'-6" to 20'-0" 15'-0" to 19'-6" 9'-11" to 20'-8" Manholes-overall dimensions-

several sizes used for

Duct banks 3.8-34 3.8-35 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-15 Revision:

Sheet: 8 3 of 3 Foundation Building Structure Length Width Thickness Comments Length Width Height Comments Figures Service Water Access Vault 39'-0" 15'-0" 1'-6" 38'-0" 14'-0" 14'-0" 3.8-37 Service Water Cooling Tower including Switchgear Room 311'-9" 61' 4' 307' 56' 85'-6"1.2-56 Service and Circulating Water Pumphouse C 74'-6" B 39'-4" 122'-0" 91'-2" 54'-0" 110'-8" 6' 3'

6' SW PH Stepped fdn.

SW PH Elec. Equip.

CW PH 78'-0" 37'-4" 122'-0" 118'-8" 52'-0" 118'-8" 92'-0" 19'-6" 92'-0" SW PH SW PH Elec.

Equip. Rm CW PH 1.2-46 through 1.2-48 Tank Farm. Area (Tunnels)

including Dikes and Foundations for Refueling Water Storage Tank and

Reactor Makeup Water Storage Tank 156' A 65'-10"

4'-5" 20' Overall dimensions

Under pipe chase Under Tanks 152'

63' 67' 65'-10" 92'

58'-4" 50'-10" Overall dimensions 10' 22' 1.2-23 through 1.2-25, 1.2-27 MUW Tank Dike RWS Tank Dike Waste Processing Building 188'-9" 29'-9' 86'-3" 23'-6" 4' 28' Stepped fdn. fill concrete 188'-9" 29'-9" 86'-3" 23'-6" 89' to 142', 61' Main structure Decontaminatio n room (In main structure, foundation and roof are irregular) 1.2-22 through 1.2-30 A - Approximate dimension B - Outer wall is thickened to provide full building width (78'-0") at grade. C - Circulating Water Pumphouse is attached to flumes and intake and discharge transition structures.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.8-16 Revision:

Sheet: 8 1 of 1 TABLE 3.8-16 C ATEGORY I, S TRUCTURES OTHER T HAN R EACTOR CONTAINMENT S TRUCTURE O R I TS I NTERNALS B ASIC L OAD C OMBINATIONS A ND L OAD F ACTORS Loading(1), (4)

All Structures Concrete Only Certain Structures, Where Appropriate Material Pipe Break Loads (Rr) Design Conditions

Required Strength Dead Load and Hydrostatic Load Live Load Operational Temperature Operating Basis Earthquake Safe Shutdown Earthquake Wind Tornado Wind

Lateral Earth Pressuress

Earth Pressure due to OBE Earth Pressure due to SSE Operational Piping Loads Accident Pressure Accident Piping Loads Jet Impingement Loads Missile Impact Loads Jet Force Reaction Internal Missile Loads Accident Temperature Design basis Flood

Unusual Snow Load

Stress Limit (5)

or

Design Criteria Loading Notations D L T o E o E P R s W W t E H e H s R o a R a R rj rm R rr M T a F Ls 1.0 1.0 - - - - - - - - - - - - - - - - 1.0 1.0 - 1.0 - - - - - - - - - - - - - - 1.0 1.0 - - - 1.0 - - - - - - - - - - - - 0.67 0.67 0.67 - - - - - - - 0.67 - - - - - - - 0.67 0.67 0.67 0.67 - - - - - - 0.67 - - - - - - -

Structural Steel S 0.67 0.67 0.67 - - 0.67 - - - - 0.67 - - - - - - - F s Per AISC 1.4 1.7 - - - - - 1.7 - - - - - - - - - - 1.4 1.7 - 1.9 - - - 1.7 1.9 - - - - - - - - - 1.4 1.7 - - - 1.7 - 1.7 - - - - - - - - - - 1.05 1.28 1.28 - - - - 1.28 - - 1.28 - - - - - - - 1.05 1.28 1.28 1.43 - - - 1.28 1.43 - 1.28 - - - - - - - 1.05 1.28 1.28 - - 1.3 - 1.28 - - 1.28 - - - - - - - 1.2 - - - - 1.7 - 1.7 - - - - - - - - - Normal Load

Concrete U 1.2 - - 1.9 - - - 1.7 1.9 - - - - - - - - ACI 318-71 0.63 0.63 0.63 - 0.63 - - - - - 0.63 - - - - - - - 0.63 0.63 0.63 - - - 0.63 - - - 0.63 - - - - - - - 0.63 0.63 - - - - - - - - - 0.63 0.63 - - - - 0.63 0.63 0.63 - 0.63 - - - - - - - 0.63 0.63 0.63 0.63 0.63 0.63 0.63 Elastic S

(2)

(2) 0.59 0.59 - - 0.59 - - - - - - 0.59 0.59 0.59 0.29 0.29 0.29 0.59 F s Per AISC 1.1 1.1 1.1 - 1.1 -- - - - - 1.1 - - - - - - - 1.1 1.1 1.1 - - - 1.1 - - - 1.1 - - - - - - - 1.1 1.1 - - - - - - - - - 1.7 1.1 1.1 1.1 1.1 1.1 1.1 1.1 1.1 - 1.4 - - - - - - - 1.4 1.1 - - - - 1.1

Structural Steel

Plastic Y

(*)

(*) 1.1 1.1 - - 1.1 - - - - - - 1.1 1.1 1.1 1.1 1.1 1.1 1.1 AISC, Part II 1.0 1.0 1.0 - 1.0 - - 1.0 - 1.0 1.0 - - - - - - - 1.0 1.0 1.0 - - - 1.0 1.0 - - 1.0 - - - - - - - 1.0 1.0 - - - - - 1.0 - - - 1.5 1.0 - - - 1.1 1.0 1.0 1.0 - 1.25 - - - 1.0 1.25 - - 1.25 1.0 1.0 1.0 1.0 - 1.0 1.0 1.0 - - 1.0 - - 1.0 - 1.0 - 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 - - 1.0 - - - - 1.0 - - - - - - - (3) 1.0 Unusual Load Concrete

U 1.0 1.0 1.0 - - 1.0 - - - - 1.0 - - - - - - - 1.0 1.0 ACI 318-71 (1) In above load combinations, the peak values of P a , T a , R a , R rj , R rr , R rm and M shall be combined (when acting concurrently) unless time history analysis is performed to justify otherwise. (2) Elastic cases to be check ed for overall stability by the plastic load combination cases as indicated by (*). (3) For design bases flood load case, elevation shall be the effective maximum ground water elevation i.e., El. + 20-0. (4) See Subsection 3.8.4.3 for discussion of loadings. (5) (F s = Allowable Stress)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.8-17 Revision:

Sheet: 8 1 of 1 TABLE 3.8-17 COMPUTER P ROGRAMS U SED I N THE A NALYSIS A ND DESIGN O F OTHER S EISMIC C ATEGORY I S TRUCTURES Computer Program Structures On Which Used Control & Diesel Generator Building Fuel Storage Building Main Steam and Feedwater Pipe Chase (East) Main Steam and Feedwater Pipe Chase (West)

Pre-Action Valve Area Primary Auxiliary Building Including Residual Heat Removal Equipment Vault Service Water Cooling Towers Including Switchgear Room 1. MRI/STARDYNE (Static Analysis) Service Water Pumphouse 2. MARC-CDC (Static Analysis) Containment Enclosure Building Containment Enclosure Building 3. LESCAL (Design of Reinforcing

Steel) Main Steam and Feedwater Pipe Chase (East) Containment Enclosure Ventilation Area Emergency Feedwater Pump Building Including Electrical Cable Tunnels and Penetration Areas 4. GENSAP (Static Analysis) Piping Tunnels Service Water Cooling Towers 5. MULTISPAN (Static Analysis)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3. 9(B)-1 Revision:

Sheet: 8 1 of 3 TABLE 3. 9(B)-1 SYSTEMS REQUIRING PREOPERATIONAL VIBRATION TESTING Flow Modes for Preoperational Vibration Testing System Steady State Transient Reactor Coolant Single and multiple pump Pump(s) starts and stops operation Pressurizer PORV discharge Residual Heat Removal Shutdown cooling mode None Provide letdown when RCS pressure is low Low head injection through RCS cold legs Safety Injection Intermediate head injection None Chemical and Volume Letdown flow modes None Control Provide seal water injection Normal purification High head injection Primary Component Loop A pump(s) operating None Cooling Loop B pump(s) operating Post LOCA recirculation S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3. 9(B)-1 Revision:

Sheet: 8 2 of 3 Spent Fuel Pool Pump(s) A and B operating None Cooling Service Water Loop A pumps operating None Loop B pumps operating Flow Modes for Preoperational Vibration Testing System Steady State Transient Steam Generator Flow at normal rate Open/close containment Blowdown isolation valve(s) Flow at maximum rate Condensate Two pumps operating None Feedwater Feedwater pump(s) operating SGFP trip FW pump recirculation Containment isolation valve close Emergency Feedwater EFW pump operating None Main Steam 100% power operation Turbine trip and dump and ARV Emergency feedwater pump trip (Terry turbine) Main steam headers to condenser dump lines Containment Building Pump(s) A & B operating None Spray in minimum recirculation mode back to RWST S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3. 9(B)-1 Revision:

Sheet: 8 3 of 3 Diesel Generator Fuel oil transfer pump(s)

None operating Starting air compressors operating Cooling water system operating Backup control air compressor operating

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9 B-2 Revision:

Sheet: 8 1 of 2 TABLE 3.9B-2 S YSTEMS R EQUIRING MONITORING O F T HERMAL EXPANSION D URING S TARTUP F UNCTIONAL TESTING (NORMAL O PERATING T EMPERATURE F OR THE FOLLOWING S YSTEMS I S 200 F O R HIGHER) SYSTEM CLASS LOCATION/COMMENTS Reactor Coolant (RC) 1 All Piping Within the RCS Pressure Boundary Reactor Heat (RH) 1 All Piping Within the RCS Pressure Boundary Safety Injection (SI) 1 All Piping Within the RCS Pressure Boundary Chemical & Volume Control (CS) 1 All Piping Within the RCS Pressure Boundary Reactor Coolant (RC) 2 RCP Suction to REG. HX (CS-E-2) RC Hot Leg 1 to RH-P-8A Suction RC Hot Leg 4 to RH-P-8B Suction Residual Heat (RH) 2 RH-P-8A Discharge to RH-E-9A RH-P-8B Discharge to RH-E-9B RH-E-9A Outlet to RC Cold Leg 1 RH-E-9B Outlet to RC Cold Leg 4 Chemical & Volume Control (CS) 2 REG. HX (CS-E-2) Outlet to LTDN HX (CS-E-4) REG. HX (CS-E-2) Outlet to RC Cold Loop 1&4 RCP Disch. to EX LTDN HX (CS-E-3) Main Steam (MS) 2 Stm. Gen. (RC-E-11A,B,C&D) to Class Break 2 N Feedwater (FW) 2 Stm. Gen. (RC-E-11A,B,C&D) to Class Break 2 N Steam Blowdown (SB) 2 Stm. Gen. (RC-E-11A,B,C&D) to SB-V9,10,11&12 Main Steam (MS) N High Press. & Low Press. Steam to Moisture Separators &

Turbines Feedwater (FW) N Feedwater Pumps (FW-P-32A,B)Discharge to Class Break 2 N Steam Generator Blowdown to Stm. Gen. & Wet Lay-Up Pump Stm. Gen. & Wet Lay-up Pump Disch. to Feedwater Pump (FW-P-32A,B) Discharge Side S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9 B-2 Revision:

Sheet: 8 2 of 2 SYSTEM CLASS LOCATION/COMMENTS Condensate (CO) N Heaters CO-E-22A, B&C to Heaters CO-E-23A, B&C: CO-E-24A,B&C; CO-E-25A&B Heaters CO-E-25A&B to Feedwater Pumps (FW-P-32A,B) Steam Extraction (EX) N High & Low Press. Extraction Steam to Heaters CO-E-23A,B&C; CO-E-24A,B&C; CO-E-25A&B; FW-E-26A&B Heater Drains (HD) N Heaters CO-E-24A,B&C; CO-E23A,B&C to Heaters CO-E-22A,B&C Heaters CO-E-25A&B; FW-E-26A&B to Heater Drain Tank (TK-22) Heater Drain Tank (HD-TK-22) to Heater Drain Pumps (HD-P31A&B) Suction Heater Drain Pump(HD-P31A&B) Discharge to Steam Generator Feed Pumps (FW-P-32A&B) Suction Steam Blowdown (SB) SB-V9,10,11&12 to Flash Tank (SB-TK-40)

NOTE: Additional systems such as those listed below may be monitored or visual inspected for any constraints during startup & testing: Main Steam Drains (MSD) Auxiliary Steam (AS) Auxiliary Steam Condensate (ASC) Moisture Separator Drains & Vents (MD) Heater Misc. Vents & Drains (MVD) Turbine Steam Seal System (TSS)

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-3 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-3 DESIGN LOADING COMBINATIONS FOR ASME CODE CLASS 1, 2 AND 3 COMPONENTS AND SUPPORTS (OTHER THAN PIPING SYSTEMS)

Plant Loading Conditions Design Loading Combinations Normal P + D + L + M + NOL + T Upset P + D + L + M + NOL + OBE + H + F t + T Emergency P + D + L + M + NEL + H + F t Faulted P + D + L + M + NFL + SSE + H + F t Where:

    • P = Pressure corresponding to the loading condition D = Dead weight L = Live weight of fluid handled; for liquid content of vessel or tank M = Snow or wind load for outdoor storage tank NOL = Nozzle load for Normal/Upset condition NEL = Nozzle load for Emergency condition NFL = Nozzle load for Faulted condition OBE = Operating Basis Earthquake (inertia load)

SSE = Safe Shutdown Earthquake (inertia load)

H = Dynamic fluid head e ffects (where applicable)

F t = Valve thrust loads (where applicable) *T = Thermal load (where applicable)

  • Temperature for Class 2 and 3 is used to determine allowable stress only.
    • Pressure for Class 1 includes LOCA effects.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-4 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-4 S TRESS L IMITS F OR N ONACTIVE C ATEGORY I, A SME CODE C LASS 2 A ND 3 P UMPS Plant Loading Conditions (1) Stress Limits Normal ASME III, NC-3400 or ND-3400 P m 1.1S Upset (P m or P 1) + P b 1.65S P m 1.5S Emergency (P m or P 1) + P b 1.8S P m 2.0S Faulted (P m or P 1) + P b 2.4S Where:

S = Material allowable stress at maximum temperature from Appendix I of ASME Section III P m = Primary general membrane stress, the average primary stress across the solid section under consideration.

Excludes effects of discontinuities and concentrations. Produced by pressure and mechanical loads.

P b = Primary bending stress. This stress is produced by pressure and mechanical loads including inertia earthquake effects but excl uding effects of discontinu ities and concentrations.

P 1 = Primary local membrane stress, the average stress across any solid section under consideration. Same as P m except that discontinuities are considered.

NOTES (1) Plant loading conditions are defined in Table 3.9(B)-3.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-5 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-5 S TRESS L IMITS F OR N ONACTIVE C ATEGORY I, A SME CODE C LASS 2 A ND 3 V ALVES Plant Loading Conditions (1) Stress Limits (2)(3) Normal ASME III, NC-3500 or ND-3500 P m < 1.1S Upset (P m or P 1) + P b < 1.65S P m < 1.5S Emergency (P m or P 1) + P b < 1.8S P m < 2.0S Faulted (P m or P 1) + P b < 2.4S NOTES (1) Plant loading conditions are defined in Table 3.9(B)-3.

(2) S, P m , P 1 and defined in Table 3.9(B)-4.

(3) Reference Code Case 1635-1 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-6 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-6 S TRESS L IMITS F OR N ONACTIVE C ATEGORY I A SME CODE C LASS 1 V ALVES Plant Loading Stress Limits Condition (1)

(2),(3),(4),(5),(6)

Normal ASME III, Subsection NB-3500 P m < 1.1 S m Upset (P m or P 1) + P b < 1.65 S m S n < 3 S m P m < 1.5 S m Emergency (P m or P 1) + P b < 1.8 S m P m < 2.0 S m Faulted (P m or P 1) + P b < 2.4 S m NOTES: (1) Plant loading conditions are defined in Table 3.9(B)-3. Pressure loading includes LOCA effects.

(2) P 1 , P m and P b are defined in Table 3.9(B)-4.

(3) S m = Material stress intensity at maximum temperature form Appendix I of ASME III.

(4) Stress in the valve resulting form connecting pipe nozzle loads and internal pressure induced stresses should not exceed the limits listed in this table.

(5) Reference ASME Section III, Subsection NB-3221.

(6) Requirements of ASME III, Subsection NB-3540 must be met. Code Ca se 1552 is an acceptable alternate.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-7 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-7 S TRESS L IMITS F OR C ATEGORY I A SME CODE C LASS 2 A ND 3 V ESSELS A ND T ANKS. Plant Loading Conditions (1) Stress Limits (2) Normal ASME NC-3300, ND-3300 NC-3800, ND-3800 P m < 1.1S Upset (P m or P 1) + P b < 1.65S P m < 1.5S Emergency (P m or P 1) + P b < 1.8S P m < 2.0S Faulted (P m or P 1) + P b < 2.4S . Notes (1) Plant loading conditions are defined in Table 3.9(B)-3.

(2) S, P m , P 1 and P b are defined in Table 3.9(B)-4.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-8 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-8 A SME SECTION I II C LASS 1 PIPING S YSTEMS L OAD C OMBINATIONS A ND S TRESS L IMITS (1) Condition Load Combination Stress Category Stress Limits Stress Combination P P m S m NB-3640 DESIGN P + D + OBE P L + P b 1.5 S m EQ(9) NB-3650 P MAX+T+TR+OBE+SAD(OBE)+

Q P L+P b+P e+Q 3 S m EQ(10) NB-3650 OR T P e 3 S m EQ(12) NB-3650 P MAX + TR + OBE + Q P L+P b+Q 3 S m EQ(13) NB-3650 AND PLANT NORMAL AND

UPSET U ---- 1.0 NB-3222.4 P MAX P m 1.5xP(DESIGN) NB-3655 PLANT EMERGENCY P MAX + D + TR + DSL P L + P b 2.25 S m EQ(9) NB-3655 P MAX P m 2 x P(DESIGN) NB-3656 PLANT FAULTED P MAX+D+TR+SSE+SAD+(SSE)+

DSL+ LOCA DISP.

P L + P b 3 S m EQ(9) NB-3656 P t + D t P m + P b 0.9 Sy NB-3226 TEST PAD + D P b 1.35 Sy NB-3226 (1) Terminology and notations are defined in Table 3.9(B)-11

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9 B-9 Revision:

Sheet: 8 1 of 3 TABLE 3.9B

-9 ASME SECTION I II CLASS 2 A ND 3 ESSENTIAL PIPING SYSTEMS LOAD COMBINATIONS A ND STRESS LIMITS (1) Condition Load Combination Stress Combination Stress Limits DESIGN P NC-3640 S h PLANT NORMAL P + D T EQ(8) NC-3650 EQ(10) NC-3650 S h S A OR P + D P + D + T EQ(8) NC-3650 EQ(11) NC-3650 S h S h + S A

PLANT UPSET PMAX+D+TR+OBE+SAD(OBE)

T EQ(9) NC-3650 EQ(10) NC-3650 1.2 S h S A OR PMAX+D+TR+OBE+SAD(OBE)

PMAX+ D + T EQ(9)

EQ(11) 1.2 S h S h + S A OR PMAX + D + TR + OBE T + SAD(OBE)

EQ(9) NC-3650 EQ(10) NC-3650 1.2 S h S A OR PMAX + D + TR + OBE PMAX + D + T + SAD(OBE)

EQ(9) NC-3650 EQ(11) NC-3650 1.2 S h S h + S A AND PMAX + D EQ(8) NC-3650 S h SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9 B-9 Revision:

Sheet: 8 2 of 3 Condition Load Combination Stress Combination Stress Limits PLANT EMERGENCY PMAX + D + TR + DSL T EQ(9) NC-3650 EQ(11) NC-3650 1.2 S h S A OR PMAX + D + TR + DSL PMAX + D + T EQ(9) NC-3650 EQ(11) NC-3650 1.2 S h S h + S A AND PMAX + D EQ(8) NC-3650 S h

PLANT FAULTED PMAX CC-1606 2 x P(DESIGN)

PMAX+D+TR+SSE+SAD(SSE)+ PAD+TAD+DSL T EQ(9) NC-3650 EQ(10) NC-3650 1.8 S h S A OR PMAX+D+TR+SSE+SAD(SSE)+

PAD+TAD+DSL PMAX + D + T EQ(9) NC-3650 EQ(11) NC-3650 1.8 S h S h + S A OR PMAX+D+TR+SSE+DSL T+SAD(SSE)+PAD+TAD EQ(9) NC-3650 EQ(10) NC-3650 1.8 S h S A OR SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9 B-9 Revision:

Sheet: 8 3 of 3 Condition Load Combination Stress Combination Stress Limits PMAX+D+TR+SSE+DSL PMAX+D+T+SAD(SSE)+PAD+TAD EQ(9) NC-3650 EQ(11) NC-3650 1.8 S h S h + S A AND PMAX + D EQ(8) NC-3650 S h TEST P t + D t PAD + D Adopted From NB

-3226 0.9 Sy 1.35 Sy (1) Terminology and notations are defined in Table 3.9(B)

-11.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-10 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-10 A SME SECTION I II C LASS 2 A ND 3 N ONESSENTIAL PIPING S YSTEMS (1) Condition Load Combination Stress Combination Stress Limits DESIGN P NC-3640 S h P + D T EQ(8) NC-3650 EQ(10) NC-3650 S h S A OR

PLANT NORMAL P + D P + D + T EQ(8) NC-3650 EQ(11) NC-3650 S h S h + S A P MAX+D+TR+OBE+SAD(OBE) T EQ(9) NC-3650 EQ(10) NC-3650 1.2 S h S A OR P MAX + D + TR + OBE T + SAD(OBE) EQ(9) NC-3650 EQ(10) NC-3650 1.2 S h S A OR P MAX + D + TR + OBE P MAX + D + T + SAD(OBE) EQ(9) NC-3650 EQ(11) NC-3650 1.2 S h S h + S A AND

PLANT UPSET P MAX + D EQ(8) NC-3650 S h PLANT EMERGENCY P + D + TR + DSL EQ(9) NC-3650 1.8 S h PLANT FAULTED P + D + TR + SSE + DSL EQ(9) NC-3650 CC-1606 2.4 S h TEST P t + D t PAD + D Adopted Form NB-3226 0.9 Sy 1.35 Sy (1) Terminology and notations ar e defined in Table 3.9(B)-11

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9 B-11 Revision:

Sheet: 8 1 of 2 TABLE 3.9B-11 TERMINOLOGY A ND N OTATIONS U SED I N T ABLES 3.9(B)-8, 3.9(B)-9 A ND 3.9(B)-10 Symbols for Stress Classification and Stress Limits are in accordance with ASME Section III. Other load symbols and definitions are specified below: P - Internal design pressure

P MAX - Peak pressure, considered as a set pressure of over-pressure safety devices P t - Test pressure D - Deadweight, consisting of the weight of the pipe and pipe supported elements such as valves and flanges, including weight of insulation and contained fluid D t - Same as 'D' where pipe contents are fluid during pressure test T - Thermal loads due to:

a. Piping thermal expansion when subjected to maximum temperature difference between the fluid and the surrounding environment in the specified plant conditions, and b. Anchor displacement due to thermal movements of piping anchors TR - Thrust or transient due to safety valve discharge, valve trip or fluid flow SAD - Seismic anchor displacement (OBE or SSE), affects piping supported from different structures of relative seismic motions PAD - Anchor displacement due to pressure, e.g., containment building penetrations due to internal pressure during test or LOCA TAD - Anchor displacement due to therma l growth of the structure e.g., radial and vertical growth of Containment Building during LOCA +/-MSL- DSL - Dynamic System Load, Accident Load affecting piping as follows:

Impact from missiles or pipe whip Jet impingement External pressure LOCA - Anchor displacement due to movement of primary or secondary loop DISP. during LOCA S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9 B-11 Revision:

Sheet: 8 2 of 2 Q - Temperature gradient loads, T 1 , T 2 and (a T a - b T b) U - Cumulative usage factor OBE - Loads generated by the Operating Basis Earthquake (OBE), which is the earthquake that could reasonably be expected to affect the plant site during the operating life of the plant and which produces the vibratory ground motion for which thos e features of the nuclear plant necessary for c ontinued operation without undue risk to the health and safety of the public have been designed to remain functional. SSE - Loads generated by the Safe Shutdown Earthquake (SSE) which is the earthquake that produces the maximum vibratory ground motion for which certain structures, systems, and components important to safety and required for safe shutdown on the plant have been designed to remain functional.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-12 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-12 S TRESS L IMITS F OR A CTIVE C ATEGORY I A SME CODE C LASS 2 A ND 3 P UMPS Plant Loading Condition (1)(3) Stress Limits (2)(4) Normal ASME Section III, Subsections NC-3400 or ND-3400 Upset P m <1.0S (P m or P 1) + P b <1.5S Emergency P m <1.0S (P m or P 1) + P b <1.5S Faulted P m <1.0S (P m or P 1) + P b <1.5S NOTES (1) Plant loading conditions are defined in Table 3.9(B)-3.

(2) S, P 1 , P m and P b are defined in Table 3.9(B)-4.

(3) Identification of the specifi c transients or events to be considered under each plant condition are addressed in Regulatory Guide 1.48.

(4) For pump supports, the allowable stresses defined in AISC "Manual of Steel Construction" is used for plant condition associated with 0.5SSE. For plant conditions associated with SSE, the stresses are limited to 90 percent of yield stress for the material involved.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-13 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-13 S TRESS L IMITS F OR A CTIVE C ATEGORY I A SME CODE C LASS 2 A ND 3 V ALVES Plant Loading Condition (1)(3) Stress Limits (2)(3)(4) Normal ASME Section III, Subsections NC-3500 or ND-3500 Upset P m < 1.0S (P m or P 1) + P b < 1.5S Emergency P m < 1.0S (P m or P 1) + P b < 1.5S Faulted P m < 1.0S (P m or P 1) + P b < 1.5S NOTES (1) Plant loading conditions are defined in Table 3.9(B)-3.

(2) S, P 1 , P m and P b are defined in Table 3.9(B)-4.

(3) Stress in the valve resulting from connecting pipe nozzle loads and internal pressure induced stresses should not exceed the limits listed in this table.

(4) Reference ASME Section III, Subsection NB-3221 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-14 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-14 S TRESS L IMITS F OR A CTIVE C ATEGORY I A SME CODE C LASS 1 V ALVES Plant Loading Stress Limits Condition (1) (2),(3),(4),(5),(6)

Normal ASME III, Subsection NB-3500 Upset P m < 1.0 Sm (P m or P 1) + P b < 1.5 Sm S n < 3 Sm Emergency P m < 1.0 Sm (P m or P 1) + P b < 1.5 Sm Faulted P m < 1.0 Sm (P m or P 1) + P b < 1.5 Sm S n < 3 Sm NOTES (1) Plant loading conditions are defined in Table 3.

9(B)-3. Pressure loading includes LOCA effects.

(2) P 1 , P m and P b are defined in Table 3.9(B)-4.

(3) S m = Material stress intensity at maximum temperature form Appendix I of ASME III.

(4) Stress in the valve resulting from connecting pi pe nozzle loads and internal pressure induced stresses should not exceed the limits listed in this table.

(5) Reference ASME Section III, Subsection NB-3221.

(6) Requirements of ASME III, Subsection NB-3540 must be met. Code Case 1552 is an acceptable alternate.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-15 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-15 S TRESS A ND DEFLECTION A NALYSIS O F P RIMARY C OMPONENT COOLING W ATER P UMPS (14 X 23-S) ITEMS - SSE ANALYSIS TUAL AC E ALLOWABL Pump Casing, Primary Membrane, psi 5,746 17,500 Membrane and Bending, psi 15,715 26,250 Casing Flange, Normal Stress, psi 24,469 26,250 Casing Flange Bolts, psi 34,967 50,000 Suction Flange, Longitudinal Stress, psi 10,213 21,000 Discharge Flange, Longitudinal Stress, psi 9,093 21,000 Shaft Stress, psi 6,724 25,000 Shaft Deflection, at Coupling, in.

0.017 0.055 at Seal, in.

0.004 0.005 at Impeller, in.

0.009 0.0125 Casing Feet, Principal, psi 3,535 14,000 Casing Foot Bolts, Tension, psi 22,280 42,000 Casing Foot Shear Pins, Shear, psi 14,328 17,000 Bedplate, Principal psi 13,990 19,333 Bedplate, Side-Channels, tension, psi 9,534 14,500 Bedplate, Side-Channel Weld, Shear, psi 9,154 9,670 Anchor Bolts, Tension, psi 12,090 19,100 Shear, psi 8,910 12,800 Natural Frequency of Pump-Motor-Bedplate System, Hz 46.19 N/A S EABROOK S TATION UFSA R D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-16 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-16 S TRESS A ND DEFLECTION S UMMARY F OR CONTAINMENT S PRAY P UMPS (6 X 11 X 14 B-C D) Items - Sse Analysis Actual Allowable Natural Frequency, Hz Pump 67

- Driver 85

- Pump Casing at Suction Nozzle, psi 2,741 27,000 Shaft Stress, psi 4,486 27,000 Pedestal Weld, psi 5,350 32,400 Pump Anchor Pin, psi 10,500 32,400 Base Cross Member, psi 20,374 32,400 Weld, psi 19,200 32,400 Bearing Load (Double Row), lbs.

1,140 17,200 (Single Row), lbs.

228 7,670 Base Hold-down bolts, Tension, psi 3,450 19,100 Shear, psi 2,430 9,900 Deflections Pump shaft, in.

0.0038 0.022 Coupling parallel misalignment, in.

0.0385 0.102 Coupling angular misalignment, degrees 0.264

1.5 Motor

rotor, in.

0.0011 0.043 S EABROOK S TATION UFSA R D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-17 Revision:

Sheet: 9 1 of 1 TABLE 3.9(B)-17 STRESS AND DEFLECTION

SUMMARY

FOR EMERGENCY FEEDWATER PUMPS (4 X 9 NH-10) Items - See Analysis Actual Allowable A. Turbine-Driven Pump Natural frequency of pump-turbine, Hz Entire assembly 48.0 -

Shaft 93.2 -

Deflection, in. Shaft @ seal 0.0042 0.005 Coupling 0.0081 0.102 B. Motor-Driven Pump Natural frequency of pump-motor Entire assembly, Hz 49.3 -

Shaft, Hz 93.2 -

Pump pedestal weld, psi Top 1,900 18,000 Bottom 15,100 18,000 Deflection, in. Shaft @ seal 0.0042 0.005 Coupling 0.0081 0.102

S EABROOK S TATION UFSA R D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-18 Revision:

Sheet: 12 1 of 1 TABLE 3.9(B)-18 S TRESS A ND DEFLECTION S UMMARY F OR S ERVICE W ATER P UMP (INGERSOLL-DRESSER PUMP 42APK) Items - See Analysis Actual Allowable Natural Frequency, Hz Horizontal 9.7 -

Vertical > 29 -

Pump Casing, psi Discharge Head at Nozzle 8,046 15,700 Column at Flange 10,460 28,500 Bowl 7,635 28,350 Lineshaft, psi 9,060 21,500 Seismic Support, psi 2,448 13,345 Base Bolts, psi 16,711 25,000 Clearance, in.

Top Impeller Wear Ring 0.006 0.0145 Lower Impeller Wear Ring 0.006 0.0145 Motor Air Gap 0.001808 0.035

S EABROOK S TATION D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-19 Revision:

Sheet: 12 1 of 1 TABLE 3.9(B)-19 STRESS AND DEFLECTION

SUMMARY

FOR COOLING TOWER PUMP (INGERSOLL-DRESSER PUMP 29LKX)

Items - See Analysis Actual Allowable Natural Frequency, Hz Pump Assembly 14.6-Motor Rotor 49.7-Pump Casing, psi Discharge Head Flange 19,437 20,550 Column Flange 21,960 28,500 Pump Casing 9,833 28,850 Shaft, psi 13,205 21,500 Seismic Supports, psi 891 23,550 Anchor Bolts, Tension, psi 16,416 17,800 Shear, psi 4,558 10,000 Clearances, in.

Impeller to Casing 0.0 0 0.0 1 Rotor/Stator 0.01 7 0.0 3 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-20 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-20 S TRESS A ND DEFLECTION S UMMARY F OR D IESEL F UEL O IL T RANSFER P UMP (N 3 DBS - 8 I MO P UMP) ITEMS - SSE ANALYSIS ACTUAL ALLOWABLE Natural Frequency, Hz 228 - Pump Casing at Inlet Nozzle, psi 7,123 26,250 Base Plate, psi 4,147 21,750 Pump Hold-Down Bolts, psi Tension 2,300 19,100 Shear 1,150 9,900 Clearances, in. 0.00009 0.0005 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-21 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-21 S TRESS A NALYSIS S UMMARY ASME IIIC LASS 1 PIPING RCS PRESSURIZER S AFETY A ND R ELIEF V ALVES S YSTEM (P&ID RC-20846) (LINE NOS: 74, 75, 76 & 80)

EVALUATION MAX STRESS (KSI) LINE NO. & COMPONENT ALLOWABLE (KSI) Eq. 9 - Design P+D+OBE 23.2 Line 76 Elbow 1.5 Sm 24.63 Eq. 9 - Faulted

P+D+TR+SSE 39.1 Line 76 Elbow 3.0 Sm 49.26 Eq. 12 T 36.0 Line 80 Transition 3.0 Sm 49.26 Eq. 13 P+D+OBE+Q 37.9 Line 80 Transition 3.0 Sm 49.26 Fatigue Usage

Factor 0.95 Line 80 Transition 1.0

  • This table is being maintained for historical purposes only. Refer to applicable calculations for latest information.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 1 of 16 TABLE 3.9B-22 S TRESS A NALYSIS S UMMARY , ASME III C LASS 2 A ND 3 PIPING (E QUATION 9) NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI)

  • This table is being maintained for historical purposes only. Refer to applicable calculations for latest information.

REACTOR COOLANT SYSTEM 3"x4"- RC-V24 Line No. 14-1

SC 2 RC-20841 16.36 19.44 9.09 19.44 26.68 29.16 3"x4"- RC-V89 Line No. 88-1

SC 2 RC-20844 18.27 19.44 8.64 19.44 27.91 29.16 REFUELING CAVITY CLEANUP SYSTEM 3/4"x1" - SF-V101 Line No. 1743-9

SC 2 SF-20484 13.24 19.08 1.17 19.08 21.66 28.62 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 2 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) FLOOR & EQUIPMENT DRAIN SYSTEM 3/4"x1"- WLD-V209 Line No. 2076-11

SC 2 WLD-20219 10.5 22.0 2.3 22.0 21.81 33.0 1"x3/4" - RH-V13 Line No. 172-1

SC 2 RH-20662 11.41 19.44 1.97 19.44 20.87 29.16 1"x3/4" - RH-V25 Line No. 169-1

SC 2 RH-20663 13.18 19.44 2.73 19.44 23.12 29.16 3/4"x1" - SI-V247 Line No. 248-16

SC 2 SI-20450 13.79 19.08 4.40 19.08 23.19 28.62 1"x2" - SI-V10 Line No. 220-1

SC 2 SI-20450 9.58 19.92 7.56 19.92 11.60 29.88 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 3 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 1"x2" - SI-V30 Line No. 221-1

SC 2 SI-20450 9.77 19.92 4.91 19.92 14.62 29.88 1"x2" - SI-V45 Line No. 222-1

SC 2 SI-20450 9.77 19.92 4.91 19.92 14.62 29.88 1"x2" - SI-V60 Line No. 223-1

SC 2 SI-20450 9.77 19.92 4.91 19.92 14.69 29.88 3/4"x1" - SI-V101 Line No. 252-1

SC 2 SI-20446 15.23 19.92 4.08 19.92 26.34 29.88 3/4"x1" - SI-V113 Line No. 253-1

SC 2 SI-20446 14.88 19.92 3.54 19.92 26.06 29.88 3/4"x1" - SI-V76 Line No. 254-1

SC 2 SI-20447 10.61 19.92 4.41 19.92 17.25 29.88 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 4 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 3/4"x1" - SI-V175 Line No. 278-2

SC 2 SI-20447 15.43 22.56 4.48 22.56 26.38 33.84 3/4"x1" - SI-V248 Line No. 255-10

SC 2 SI-20446 4.34 19.68 3.38 19.68 5.31 29.52 3/4"x1" - CBS-V62 Line No. 1228-1

SC 2 SI-20446 8.74 19.92 2.02 19.92 15.46 29.88 CHEMICAL & VOLUME CONTROL SYSTEM 3"x2" - CS-V148 Line No. 353-1

SC 2 CS-20722 2.12 20.64 1.68 20.64 6.35 30.96 3/4"x1" - CS-V794 Line No. 332-26

SC 2 CS-20726 0.82 20.64 0.35 20.64 1.29 30.96 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 5 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 2"x3" - CS-V173 Line No. 353-1

SC 2 CS-20726 2.12 20.64 1.68 20.64 2.55 30.96 4"x3" - CS-V243 Line No. 384

SC 2 CS-20725 3.11 19.54 1.88 19.54 4.94 29.30 1"x3/4" - CS-V492 Line No. 372-1

SC 2 CS-20725 17.59 22.06 6.03 22.06 24.53 33.09 1"x3/4" - CS-V227 Line No. 373-1

SC 2 CS-20725 6.21 19.92 1.08 19.92 11.33 29.88 3"x2" - CS-V250 Line N-378-1

SC 2 CS-20726 8.10 19.54 4.43 19.54 11.77 29.30 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 6 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) SPENT FUEL POOL COOLING AND CLEANUP SYSTEM 1"x3/4" - SF-V74 Line No. 1741-1

SC 3 SF-20482 3.14 19.92 1.58 19.92 4.7 29.88 1"x3/4" - SF-V45 Line No. 1723-1

SC 3 SF-20482 3.14 19.92 1.58 19.92 4.7 29.88 1"x3/4" - SF-V183 Line No. 1766-1

SC 3 SF-20482 4.04 21.19 2.56 21.19 6.75 31.78 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 7 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) PRIMARY COMPONENT COOLING WATER SYSTEM 3/4"x1" - CC-V320 Line No. 807-2

SC 3 CC-20211 0.97 18.00 0.35 18.00 1.60 27.00 3/4"x1" - CC-V321 Line No. 809-5

SC 3 CC-20211 8.45 18.00 0.35 18.00 16.55 27.00 3"x4" - CC-V486 Line No. 777-31

SC 3 CC-20207 14.47 18.00 2.11 18.00 22.94 27.00 11/2"x21/2" - CC-V326 Line No. 714-4

SC 3 CC-20209 6.15 18.00 4.92 18.00 7.37 27.00 11/2"x2" - CC-V364 Line No. 713-7

SC 3 CC-20207 5.87 18.00 4.00 18.00 7.75 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 8 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 11/2"x21/2" - CC-V242 Line No. 718-4

SC 3 CC-20209 5.65 18.00 5.11 18.00 6.19 27.00 11/2"x2" - CC-V840 Line No. 798-45

SC 2 CC-20213 9.61 18.00 2.51 18.00 16.71 27.00 11/2"x2" - CC-V474 Line No. 827-20

SC 2 CC-20213 11.19 18.00 2.31 18.00 20.08 27.00 3"x4" - CC-V120 Line No. 827-27

SC 3 CC-20214 14.73 18.00 2.14 18.00 23.28 27.00 11/2"x21/2" - CC-V435 Line No. 712-4

SC 3 CC-20209 14.62 18.00 6.00 18.00 23.23 27.00 11/2"x2" - CC-V363 Line No. 785-7

SC 3 CC-20209 4.04 18.00 1.79 18.00 6.30 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 9 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 11/2x2" - CC-V235 Line No. 709-7

SC 2 CC-20215 3.68 18.00 1.66 18.00 5.70 27.00 11/2"x2" CC-V408 Line No. 790-9

SC 3 CC-20209 7.86 18.00 6.33 18.00 9.39 27.00 3/4"x1" - CC-V30 Line No. 758-5

SC 3 CC-20205 8.45 18.00 0.35 18.00 16.55 27.00 2"x3" - CC-V343 Line No. 816-5

SC 3 CC-20206 0.98 18.00 0.77 18.00 1.25 27.00 3/4"x1" - CC-V342 Line No. 815-3

SC 3 CC-20206 2.68 18.00 1.63 18.00 3.73 27.00 11/2"x2" - CC-V845 Line No. 752-25

SC 2 CC-20207 10.91 18.00 2.31 18.00 19.51 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 10 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 11/2"x2" - CC-V410 Line No. 777-13

SC 2 CC-20207 8.57 18.00 2.31 18.00 14.84 27.00 11/2"x21/2" - CC-V407 Line No. 782-6

SC 3 CC-20207 5.21 18.00 0.66 18.00 9.78 27.00 3/4"x1" - CC-V135 Line No. 781-7

SC 3 CC-20207 15.59 18.00 5.11 18.00 26.08 27.00 3/4"x1" - CC-V141 Line No. 783-4

SC 3 CC-20207 4.40 18.00 1.52 18.00 7.27 27.00 3/4"-1" - CC-V26 Line No. 759-4

SC 3 CC-20207 1.95 18.00 0.38 18.00 3.5 27.00 3"x4" - CC-V143 Line No. 784-5

SC 3 CC-20207 0.91 18.00 0.74 18.00 1.07 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 11 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 11/2"x21/2 - CC-V409 Line No. 793-12

SC 3 CC-20205 3.72 18.00 2.07 18.00 5.37 27.00 11/2"x2" - CC-V647 Line No. 795-7

SC 3 CC-20205 16.45 18.00 3.69 18.00 19.14 27.00 11/2"X21/2" - CC-V264 Line No. 833-3

SC 3 CC-20213 6.23 18.00 0.66 18.00 11.8 27.00 3/4"-1" - CC-V262 Line No. 832-2

SC 3 CC-20213 4.04 18.00 1.85 18.00 6.23 27.00 3/4"-1" - CC-V269 Line No. 834-4

SC 3 CC-20213 3.31 18.00 1.05 18.00 5.58 27.00 3/4"-1" - CC-V322 Line No. 861-4

SC 3 CC-20213 1.93 18.00 0.38 18.00 3.49 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 12 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 3/4"-4" - CC-V271 Line No. 835-5

SC 3 CC-20213 0.96 18.00 0.74 18.00 1.17 27.00 11/2"x21/2" - CC-V442 Line No. 837-1

SC 3 CC-20211 5.27 18.00 1.94 18.00 8.61 27.00 61/2"x2" - CC-V171 Line No. 794-6

SC 3 CC-20212 14.42 18.00 3.69 18.00 22.22 27.00 SERVICE WATER SYSTEM 2"x3" - SW-V32 Line No. 1810-6

SC 3 SW-20795 9.14 18.00 2.05 18.00 16.30 27.00 3/4"x3/4" - SW-V514-A Line No. 1834-5

SC 3 SW-20795 4.75 9.72 2.13 9.72 7.38 14.58 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 13 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 3/4"x3/4" - SW-V514-B Line No. 1836-5

SC 3 SW-20795 8.45 9.72 2.59 9.72 14.3 14.58 2"x3" - SW-V73 Line No. 1812-6

SC 3 SW-20795 6.28 18.00 1.42 18.00 11.14 27.00 11/2"x2" - SW-V214 Line No. 1846-7

SC 3 SW-20794 2.28 18.00 0.78 18.00 3.79 27.00 DEMINERALIZED WATER SYSTEM 11/2"x2" - DM-V18 Line No. 1608-6

SC 2 DM-20352 4.74 19.78 1.15 19.78 16.45 29.67 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 14 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) REACTOR COOLANT DRAIN TANK 11/2"x2" - WLD-V213 Line No. 2102-12

SC 2 WLD-20218 12.26 19.68 3.17 19.68 21.35 29.52 MAIN STEAM 4000-41 (30") MS-20580 19.88 21.00 10.91 21.00 25.36 31.50 4000-41 (30") MS-20583 19.88 21.00 10.91 21.00 25.36 31.50 4001-41 (30") MS-20581 19.88 21.00 10.91 21.00 25.36 31.50 4001-41 (30") MS-20583 19.88 21.00 10.91 21.00 25.36 31.50 4002-37 (30") MS-20581 19.88 21.00 10.91 21.00 25.36 31.50 4002-37 (30") MS-20583 19.88 21.00 10.91 21.00 25.36 31.50 4003-37 (30") MS-20580 19.88 21.00 10.91 21.00 25.36 31.50 4003-37 (30") MS-20583 19.88 21.00 10.91 21.00 25.36 31.50 4000-11 (6") MS-20580 11.52 18.00 7.72 18.00 20.54 27.00 4001-11 (6") MS-20581 11.52 18.00 7.72 18.00 20.54 27.00 4002-09 (6") MS-20581 11.03 18.00 7.65 18.00 17.59 27.00 4003-09 (6") MS-20580 11.03 18.00 7.65 18.00 17.59 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 15 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 4000-15 (10") MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4000-19 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4000-23 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4000-27 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4000-30 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4001-15 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4001-19 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4001-23 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4001-27 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4001-30 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4002-13 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4002-17 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4002-21 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4002-25 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4002-28 MS-20581 1.33 18.00 1.33 18.00 1.33 27.00 4003-13 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4003-17 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4003-21 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-22 Revision:

Sheet: 8 16 of 16 NORMAL & UPSET CONDITION EMERGENCY CONDITION FAULTED CONDITION PIPING SYSTEM SYSTEM DRAWING NO.

ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) ACTUAL (KSI) 4003-25 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 4003-28 MS-20580 1.33 18.00 1.33 18.00 1.33 27.00 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-23 Revision:

Sheet: 8 1 of 1 TABLE 3.9(B)-23 ASME CODE C LASS 1, 2 A ND 3 P IPE S UPPORT L OAD C OMBINATIONS Normal & Upset Faulted D, T, OBE, TR, D, T, SSE, TR, SAD (SSE), PAD, SAD (OBE) TAD, LOCA displacement Definitions of Acronyms Used Above Symbols for Stress Classification and Stress Limits are in accordance with ASME Section III. Other load symbols and definitions are specified below: D - Deadweight, consists of the weight of the pipe and pipe supported elements such as valves and flanges, including weight of insulation and contained fluid. T - Thermal loads due to:

a. Piping thermal expansion when subjected to maximum temperature difference between the fluid and the surrounding environment in the specified plant conditions, and
b. Anchor displacement due to thermal movements of piping anchors TR - Thrust or transient due to safety valve discharge, valve trip or fluid flow SAD - Seismic anchor displacement (OBE or SSE), affects piping supported from different structures of relative seismic motions PAD - Anchor displacement due to pressure, e.g., containment building penetra tions due to internal pressure during test or LOCA TAD - Anchor displacement due to thermal growth of the structure e.g., radial and vertical growth of Containment Building during LOCA +/- MSL-Dynamic System Load, Accident Load affecting piping as follows:

Impact from missiles or pipe whip Jet impingement External pressure OBE - Loads generated by the Operating Basis Earthquake (OBE), which is the earthquake that could reasonably be expected to affect the plant site during the operating life of the plant and which produces the vibratory ground motion for which those features of the nuclear plant necessary for continued operation without undue risk to the health and safety of the public have been designed to remain functional. SSE - Loads generated by the Safe Shutdown Earthquake (SSE) which is the earthquake that produces the maximum vibratory ground motion for which certain structures, systems, and components important to safety and required for safe shutdown on the plant have been designed to remain functional.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-26 Revision:

Sheet: 9 1 of 1 TABLE 3.9(B)-26 A CTIVE P UMPS Pump Item No.

System ASME Safety Class Normal Mode Post Loca Mode Function PCCW Pumps CC-11A, B, C, D CC 3 ON ON Primary Component Cooling Water SW Pumps SW-P-41A, B, C, D SW 3 ON ON Ultimate Heat Sink (Ocean) Cooling Tower Pumps SW-P-110A, B SW 3 OFF ON Ultimate Heat Sink (Tower) EFW Pumps FW-P-37A, B FW 3 OFF ON Emergency Feed Water Diesel Generator Fuel Oil Pumps

  • DG-P-38A, B DG 3 OFF ON Day Tank Fuel Supply Containment Spray Pump CBS-P-9A, B CBS 2 OFF ON Containment Building Spray Spent Fuel Pool Cooling Pumps
  • Tested periodically by Station procedures, but not included in the IST program.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 1 of 14 TABLE 3.9(B)-27 A CTIVE V ALVE L IST B Y S YSTEM SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY AS V-175 Gate 3 12.00 Motor AS V-176 Gate 3 12.00 Motor CC MM-762 Relief (1) 3 10.00 Self CC MM-763 Relief (1) 3 10.00 Self CC V-1 Check 3 24.00 Delta P CC V-4 Check 3 24.00 Delta P CC V-32 Butt 3 10.00 Air CC V-57 Butt 2 12.00 Air CC V-121 Butt 2 12.00 Air CC V-122 Butt 2 12.00 Air CC V-137 Butt 3 14.00 Motor CC V-145 Butt 3 16.00 Motor CC V-168 Butt 2 12.00 Air CC V-175 Butt 2 12.00 Air CC V-176 Butt 2 12.00 Air CC V-256 Butt 2 12.00 Air CC V-257 Butt 2 12.00 Air CC V-266 Butt 3 14.00 Motor CC V-272 Butt 3 16.00 Motor CC V-295 Check 3 24.00 Delta P (1) These devices are rupture discs and are classified as valves in the ASME Code. However, end loading calculations do not apply.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 2 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY CC V-298 Check 3 24.00 Delta P CC V-341 Butt 3 6.00 Air CC V-410 Relief 2 1.50 Self CC V-426 Butt 3 20.00 Air CC V-427 Butt 3 20.00 Air CC V-445 Butt 3 10.00 Air CC V-447 Butt 3 20.00 Air CC V-448 Butt 3 20.00 Air CC V-474 Relief 2 1.50 Self CC V-840 Relief 2 1.50 Self CC V-845 Relief 2 1.50 Self CC V-1298 Ball 3 1.00 Air CC V-1301 Ball 3 1.00 Air CC V-975 Plug 3 1.00 Air CC V-986 Plug 3 1.00 Air CC V-1105 Relief 2 0.75 Self CC V-1112 Relief 2 0.75 Self CC V-1277 Relief 3 0.75 Self CC V-1278 Relief 3 0.75 Self CC V-1279 Relief 3 0.75 Self CC V-1282 Check 3 2.00 Delta P CC V-1283 Check 3 2.00 Delta P CC TV-2171-1 Butt 3 24.00 Air CC TV-2171-2 Butt 3 24.00 Air CC TV-2271-1 Butt 3 24.00 Air CC TV-2271-2 Butt 3 24.00 Air S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 3 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY CO V-421 Check 3 2.00 Delta P CO V-422 Check 3 2.00 Delta P CS V-2 Check 1 2.00 Delta P CS V-4 Check 2 2.00 Delta P CS V-18 Check 1 2.00 Delta P CS V-20 Check 2 2.00 Delta P CS V-34 Check 1 2.00 Delta P CS V-36 Check 2 2.00 Delta P CS V-50 Check 1 2.00 Delta P CS V-52 Check 2 2.00 Delta P CS V-148 Relief 2 2.00 Self CS V-199 Check 2 2.00 Delta P CS V-211 Check 2 2.00 Delta P CS V-410 Gate 3 4.00 Manual CS V-416 Gate 3 4.00 Manual CS V-423 Weir 3 2.00 Manual CS V-424 Weir 3 2.00 Manual CS V-427 Check 1 2.00 Delta P CS V-430 Weir 3 2.00 Manual CS V-431 Weir 3 2.00 Manual CS V-442 Gate 2 4.00 Manual CS V-449 Check 3 2.00 Delta P CS V-453 Check 3 2.00 Delta P CS V-471 Check 1 2.00 Delta P CS V-472 Check 1 2.00 Delta P CS V-473 Check 1 2.00 Delta P CS V-474 Check 1 2.00 Delta P S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 4 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY CS V-794 Relief 2 0.75 Self CS V-828 (2) Gate 2 3.00 Manual CS V-829 (2) Gate 2 3.00 Manual DG A FY-AS1 (3) Three-way 3 0.375 Solen DG A FY-AS2 (3) Three-way 3 0.375 Solen DG TCV7A-1 (3) Three-way 3 6.00 Air DG TCV7A-2 (3) Three-way 3 6.00 Air DG V-2A (3) Check 3 6.00 Delta P DG V-5A (3) Check 3 6.00 Delta P DG V-23A (3) Check 3 5.00 Delta P DG V-24A (3) Check 3 5.00 Delta P DG V-29A (3) Three-way 3 5.00 Self DG V-62A Relief 3 0.75 Self DG V-66A Relief 3 0.75 Self DG V-69A Check 3 0.75 Delta P DG V-70A Check 3 0.75 Delta P DG V-84A (3) Check 3 1.00 Delta P DG B-FY-AS1 (3) Three-way 3 0.375 Solen DG B-FY-AS2 (3) Three-way 3 0.375 Solen DG TCV7B-1 (3) Three-way 3 6.00 Air DG TCV7B-2 (3) Three-way 3 6.00 Air DG V-2B (3) Check 3 6.00 Delta P DG V-5B (3) Check 3 6.00 Delta P DG V-23B (3) Check 3 5.00 Delta P DG V-24B (3) Check 3 5.00 Delta P DG V-29B (3) Three-way 3 5.00 Self (2) This valve may require repositioning only during shutdown beyond hot standby. As such, it is not required to be included in the IST Program; it may be periodically tested by another station procedure.

(3) This is a skid-mounted component that is excluded from the IST Program and is periodically tested by another Station procedure.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 5 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY DG V-62B Relief 3 0.75 Self DG V-66B Relief 3 0.75 Self DG V-69B Check 3 0.75 Delta P DG V-70B Check 3 0.75 Delta P DG V-84B (3) Check 3 1.00 Delta P DG V-115 Check 3 1.50 Delta P DG V-121 Check 3 1.50 Delta P DM V-18 Relief 2 1.50 Self DM V-611 Check 3 6.00 Delta P DM V-612 Check 3 6.00 Delta P FW V-30 Gate 2 18.00 Air FW V-39 Gate 2 18.00 Air FW V-48 Gate 2 18.00 Air FW V-57 Gate 2 18.00 Air FW V-64 Check 3 6.00 Delta P FW V-70 Check 3 6.00 Delta P FW V-76 Schk 2 4.00 Delta P FW V-82 Schk 2 4.00 Delta P FW V-88 Schk 2 4.00 Delta P FW V-94 Schk 2 4.00 Delta P FW V-216 Schk 3 6.00 Delta P FW V-330 Check 2 18.00 Delta P FW V-331 Check 2 18.00 Delta P FW V-332 Check 2 18.00 Delta P FW V-333 Check 2 18.00 Delta P (3) This is a skid-mounted component that is excluded from the IST Program and is periodically tested by another Station procedure.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 6 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY FW FV-4214-A Globe 3 4.00 Motor FW FV-4214-B Globe 3 4.00 Motor FW FV-4224-A Globe 3 4.00 Motor FW FV-4224-B Globe 3 4.00 Motor FW FV-4234-A Globe 3 4.00 Motor FW FV-4234-B Globe 3 4.00 Motor FW FV-4244-A Globe 3 4.00 Motor FW FV-4244-B Globe 3 4.00 Motor FW V-357 Check 3 6.00 Delta P FW V-346 Globe 3 4.00 Motor FW V-347 Globe 3 4.00 Motor FW V-349 Check 3 4.00 Delta P FW V-350 Check 3 3.00 Delta P FW V-353 Check 3 3.00 Delta P FW V-351 Check 3 1.00 Delta P IA V-531 Check 2 2.00 Delta P MS V-6 Relief 2 6.00 Self MS V-7 Relief 2 6.00 Self MS V-8 Relief 2 6.00 Self MS V-9 Relief 2 6.00 Self MS V-10 Relief 2 6.00 Self MS V-22 Relief 2 6.00 Self MS V-23 Relief 2 6.00 Self MS V-24 Relief 2 6.00 Self MS V-25 Relief 2 6.00 Self S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 7 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY MS V-26 Relief 2 6.00 Self MS V-36 Relief 2 6.00 Self MS V-37 Relief 2 6.00 Self MS V-38 Relief 2 6.00 Self MS V-39 Relief 2 6.00 Self MS V-40 Relief 2 6.00 Self MS V-50 Relief 2 6.00 Self MS V-51 Relief 2 6.00 Self MS V-52 Relief 2 6.00 Self MS V-53 Relief 2 6.00 Self MS V-54 Relief 2 6.00 Self MS V-86 Gate 2 30.00 Phyd MS V-88 Gate 2 30.00 Phyd MS V-90 Gate 2 30.00 Phyd MS V-92 Gate 2 30.00 Phyd MS V-94 Check 3 4.00 Delta P MS V-96 Check 3 4.00 Delta P MS V-204 Globe 2 4.00 Motor MS V-205 Globe 2 4.00 Motor MS V-206 Globe 2 4.00 Motor MS V-207 Globe 2 4.00 Motor MS PV-3001 Globe 2 10.00 Air MS PV-3002 Globe 2 10.00 Air MS PV-3003 Globe 2 10.00 Air MS PV-3004 Globe 2 10.00 Air S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 8 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY MS V-400 Check 3 0.75 Delta P MS V-401 Check 3 0.75 Delta P MS V-404 Check 3 0.75 Delta P MS V-405 Check 3 0.75 Delta P MS V-417 Check 3 0.75 Delta P MS V-418 Check 3 0.75 Delta P MS V-393 Globe 2 3.00 Air MS V-394 Globe 2 3.00 Air MS V-395 Globe 3 4.00 Air NG V-14 Globe 2 1.00 Air NG FV-4609 Globe 2 1.00 Solen NG FV-4610 Globe 2 1.00 Solen RC V-312 Relief 2 0.75 Self RC V-314 Relief 2 0.75 Self RC V-323 Globe 2 0.75 Motor RC V-337 Relief 2 0.75 Self RC V-360 Relief 2 0.75 Self RC V-361 Relief 2 0.75 Self RC V-475 Check 2 0.50 Delta P RC V-479 Check 2 0.50 Delta P RC FV-2830 Globe 2 0.50 Solen RC FV-2831 Globe 2 0.50 Solen RC FV-2832 Globe 2 0.50 Solen RC FV-2833 Globe 2 0.50 Solen RC FV-2836 Globe 2 0.50 Solen RC FV-2837 Globe 2 0.50 Solen RC FV-2840 Globe 2 0.50 Solen S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 9 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY RC FV-2874 Globe 2 0.50 Solen RC FV-2876 Globe 2 0.50 Solen RC FV-2881 Globe 2 0.75 Solen RC FV-2894 Globe 2 0.50 Solen RC FV-2896 Globe 2 0.50 Solen RH V-16 Globe 2 0.75 Air RH V-17 Globe 2 0.75 Air RH V-18 (2) Globe 2 2.00 Manual RH V-19 (2) Globe 2 2.00 Manual SB V-1 Gate 2 3.00 Air SB V-3 Gate 2 3.00 Air SB V-5 Gate 2 3.00 Air SB V-7 Gate 2 3.00 Air SB V-9 Gate 2 3.00 Air SB V-10 Gate 2 3.00 Air SB V-11 Gate 2 3.00 Air SB V-12 Gate 2 3.00 Air SF V-3 (4) Check 3 6.00 Delta P SF V-7 (4) Check 3 6.00 Delta P SF V-101 Relief 2 0.75 Self SI V-81 Check 1 2.00 Delta P SI V-86 Check 1 2.00 Delta P SI V-88 Check 2 1.50 Delta P SI V-91 Check 2 1.50 Delta P SI V-106 Check 1 2.00 Delta P SI V-110 Check 1 2.00 Delta P (2) This valve may require repositioning only during shutdown beyond hot standby. As such, it is not required to be included in the IST Program; it may be periodically tested by another station procedure.

(4) Tested periodically by Sta tion procedures, but not included in the IST Program because spent fuel pool cooling function does not meet ASME OM Code criteria.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 10 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY SI V-118 Check 1 2.00 Delta P SI V-122 Check 1 2.00 Delta P SI V-126 Check 1 2.00 Delta P SI V-130 Check 1 2.00 Delta P SI V-144 Check 1 1.50 Delta P SI V-148 Check 1 1.50 Delta P SI V-152 Check 1 1.50 Delta P SI V-156 Check 1 1.50 Delta P SI V-247 Relief 2 0.75 Self SI FV-2475 (5) Globe 2 1.00 Solen SI FV-2476 (5) Globe 2 1.00 Solen SI FV-2477 (5) Globe 2 1.00 Solen SI FV-2482 (5) Globe 2 1.00 Solen SI FV-2483 (5) Globe 2 1.00 Solen SI FV-2486 (5) Globe 2 1.00 Solen SI FV-2495 (5) Globe 2 1.00 Solen SI FV-2496 (5) Globe 2 1.00 Solen RH FCV-610 Globe 2 3.00 Motor RH FCV-611 Globe 2 3.00 Motor SS V-273 Check 2 0.50 Delta P SS FV-2857 Globe 2 0.50 Solen SW V-1 Check 3 24.00 Delta P SW V-2 Butt 3 24.00 Motor SW V-3 Check 3 24.00 Delta P (5) This valve may be required to be repositioned during a plant cooldown beyond hot standby and is required to remain in its normal position during Modes 1, 2, and 3. The normal passive function must be included in the IST Program. Its active function may be periodically tested by another Station procedure.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 11 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY SW V-4 Butt 3 12.00 Motor SW V-5 Butt 3 12.00 Motor SW V-16 Butt 3 16.00 Air SW V-18 Butt 3 16.00 Air SW V-19 Butt 3 24.00 Motor SW V-20 Butt 3 24.00 Motor SW V-22 Butt 3 24.00 Motor SW V-23 Butt 3 24.00 Motor SW V-24 Check 3 24.00 Delta P SW V-25 Butt 3 24.00 Motor SW V-27 Butt 3 24.00 Motor SW V-28 Check 3 24.00 Delta P SW V-29 Butt 3 24.00 Motor SW V-30 Check 3 24.00 Delta P SW V-31 Butt 3 24.00 Motor SW V-34 Butt 3 24.00 Motor SW V-53 Check 3 24.00 Delta P SW V-54 Butt 3 24.00 Motor SW V-56 Butt 3 24.00 Motor SW V-74 Butt 3 24.00 Motor SW V-76 Butt 3 24.00 Motor SW V-139 Butt 3 24.00 Motor SW V-140 Butt 3 24.00 Motor

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 12 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY SW V-174 Check 3 1.00 Delta P SW V-175 Check 3 1.00 Delta P SW V-176 Check 3 1.00 Delta P SW V-177 Check 3 1.00 Delta P VG FV-1661 Globe 2 2.00 Solen VG FV-1712 Globe 2 2.00 Solen CAH FV-6572 Gate 2 0.50 Solen CAH FV-6573 Gate 2 0.50 Solen CAH FV-6574 Gate 2 0.50 Solen CAH V-12 Check 2 0.50 Delta P CAP V-1 Butt 2 36.00 Air CAP V-2 Butt 2 36.00 Air CAP V-3 Butt 2 36.00 Air CAP V-4 Butt 2 36.00 Air CBS V-3 Check 2 12.00 Delta P CBS V-7 Check 2 12.00 Delta P CBS V-8 Gate 2 16.00 Motor CBS V-9 Check 2 12.00 Delta P CBS V-11 Gate 2 8.00 Motor CBS V-12 Check 2 8.00 Delta P CBS V-14 Gate 2 16.00 Motor CBS V-15 Check 2 12.00 Delta P CBS V-17 Gate 2 8.00 Motor CBS V-18 Check 2 8.00 Delta P CBS V-25 Check 2 16.00 Delta P S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 13 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY CBS V-26 Check 2 16.00 Delta P CBS V-31 Butt 2 4.00 Air CBS V-32 Butt 2 4.00 Air CBS V-33 Butt 2 4.00 Air CBS V-38 Gate 2 6.00 Motor CBS V-43 Gate 2 6.00 Motor CBS V-94 Relief 2 0.75 Self CBS V-96 Relief 2 0.75 Self CBS V-147 Check 2 16.00 Delta P CBS V-148 Check 2 16.00 Delta P CBS V-149 (6) Relief 2 0.75 Self CBS V-150 (6) Relief 2 0.75 Self CBS V-151 (6) Relief 2 0.75 Self CBS V-152 (6) Relief 2 0.75 Self CGC V-3 Globe 2 1.00 Manual CGC V-4 Check 2 1.00 Delta P CGC V-10 Globe 2 1.00 Manual CGC V-12 Globe 2 1.00 Manual CGC V-13 Globe 2 1.00 Manual CGC V-14 Globe 2 2.00 Motor CGC V-24 Globe 2 1.00 Manual CGC V-25 Check 2 1.00 Delta P CGC V-28 Globe 2 2.00 Motor CGC V-32 Globe 2 1.00 Manual CGC V-34 Globe 2 1.00 Manual CGC V-35 Globe 2 1.00 Manual (6) This valve is required to perform its active safety function only during shutdown modes beyond hot standby. It is still included in the IST Program per the requirements of the ASME OM Code.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.9(B)-27 Revision:

Sheet: 13 14 of 14 SYS TAG NO TYPE SAFETY CLASS SIZE ACTUATED BY COP V-1 Butt 2 8.00 Air COP V-2 Butt 2 8.00 Air COP V-3 Butt 2 8.00 Air COP V-4 Butt 2 8.00 Air MSD V-44 Globe 2 1.00 Motor MSD V-45 Globe 2 1.00 Motor MSD V-46 Globe 2 1.00 Motor MSD V-47 Globe 2 1.00 Motor RMW V-30 Globe 2 3.00 Air RMW V-31 Weir 3 2.00 Manual RMW V-34 Weir 3 2.00 Manual RMW V-119 Check 2 2.00 Delta P WLD V-81 Globe 2 3.00 Air WLD V-82 Globe 2 3.00 Air WLD V-209 Relief 2 0.75 Self WLD V-213 Relief 2 1.50 Self WLD FV-8330 Globe 2 2.00 Solen WLD FV-8331 Globe 2 2.00 Solen

  • LEGEND*

VALVE-TYPE BUTT - Butterfly SCHK - Stop-Check ACTUATED-BY SOLEN - Solenoid PHYD - Pneumatic-Hydraulic

DELTA P - Delta Pressure

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-28 Revision:

Sheet: 16 1 of 5 TABLE 3.9(B)

-28 VALVES THAT PERFORM A MECHANICAL MOTION TO ACCOMPLISH OR SUPPORT A SAFETY FUNCTION (NON-ASME III OR NON

-1E) Tag No. Type ANS Safety Class Actuated By Normal Position CBA-TCV-21200A Three-way 3 Motor Open CBA-TCV-21200B Three-way 3 Moto r Open CC-V-120 Relief NNS Self Closed CC-V-486 Relief NNS Self Closed CS-V-834 Relief NNS Self Closed DG-V-18A Check 3 Delta P Closed DG-V-31A Check 3 Delta P Closed DG-V-41A Gate 3 Self Open DG-V-52A Regulat or 3 Self Open DG-V-54A Relief 3 Self Closed DG-V-59A Shuttle 3 Self Closed DG-V-60A Control 3 Self Closed DG-V-72A Relief 3 Self Closed DG-V-82A Check 3 Delta P Closed DG-V-83A Check 3 Delta P Closed DG-V-85A Check 3 Delta P Closed DG-V-87A Check 3 Delta P Open DG-V-94A Check 3 Delta P Open DG-V-100A Relief 3 Self Closed DG-V-196A Relief 3 Self Closed DG-V-220A Shuttle 3 Self Closed DG-V-221A Shuttle 3 Self Closed DG-V-224A Contro l 3 Self Closed DG-V-225A Gate 3 Manual Open DG-V-253A Three-way 3 Solenoid Closed DG-V-257A Regulator 3 Self Closed SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-28 Revision:

Sheet: 16 2 of 5 Tag No. Type ANS Safety Class Actuated By Normal Position DG-V-258A Relief 3 Self Closed DG-V-259A Relief 3 Self Closed DG-V-260A Check 3 Delta P Closed DG-V-261A Check 3 Delta P Closed DG-V-269A Shuttle 3 Self Closed DG-V-279A Globe 3 Solenoid Open DG-V-280A Globe 3 Solenoid Closed DG-V-281A Check 3 Delta P Closed DG-V-282A Check 3 Delta P Closed DG-V-285A Globe 3 Solenoid Closed DG-V-286A Check 3 Delta P Closed DG-V-287A Check 3 De lta P Closed DG-V-288A Globe 3 Solenoid Closed DG-V-289A Globe 3 Solenoid Closed DG-V-325A Gate 3 Solenoid Open DG-V-331A Relief 3 Self Closed DG-V-332A Ball 3 Manual Closed DG-V-333A Ball 3 Manual Closed DG-V-334A Check 3 Delta P Closed DG-V-335A Relief 3 Self Closed DG-V-18B Check 3 Delta P Closed DG-V-31B Check 3 Delta P Closed DG-V-41B Gate 3 Self Open DG-V-52B Regulator 3 Self Open DG-V-54B Relief 3 Self Closed DG-V-59B Shuttle 3 Self Closed DG-V-60B Control 3 Self Closed DG-V-72B Relief 3 Self Closed DG-V-82B Check 3 Delta P Closed SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-28 Revision:

Sheet: 16 3 of 5 Tag No. Type ANS Safety Class Actuated By Normal Position DG-V-83B Check 3 Delta P Closed DG-V-85B Check 3 Delta P Closed DG-V-87B Check 3 Delta P Open DG-V-9 4B Check 3 Delta P Open DG-V-100B Relief 3 Self Closed DG-V-196B Relief 3 Self Closed DG-V-220B Shuttle 3 Self Closed DG-V-221B Shuttle 3 Self Closed DG-V-224B Control 3 Self Closed DG-V-225B Gate 3 Manual Open DG-V-253B Three-way 3 Solenoid Closed DG-V-257B Regulator 3 Self Closed DG-V-258B Relief 3 Self Closed DG-V-259B Relief 3 Self Closed DG-V-260B Check 3 Delta P Closed DG-V-261B Check 3 Delta P Closed DG-V-269B Shuttle 3 Self Closed DG-V-279B Globe 3 Solenoid Open DG-V-280B Globe 3 Solenoid Closed DG-V-281B Check 3 Delta P Closed DG-V-282B Check 3 Delta P Closed DG-V-285B Globe 3 Solenoid Closed DG-V-286B Check 3 Delta P Closed DG-V-287B Check 3 Delta P Closed DG-V-288B Globe 3 Solenoid Closed DG-V-289B Globe 3 Solenoid Closed DG-V-325B Gate 3 Solenoid Open DG-V-331B Relief 3 Self Closed DG-V-332B Ball 3 Manual Closed DG-V-333B Ball 3 Manual Closed DG-V-334B Check 3 Delta P Closed DG-V-335B Relief 3 Self Closed SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-28 Revision:

Sheet: 16 4 of 5 Tag No. Type ANS Safety Class Actuated By Normal Position DGA-PCV-9602-1 Regulator 3 Self Open DGA-PCV-9612-1 Regulator 3 Self Open DGB-PCV-9602-1 Regulator 3 Self Open DGB-PCV-9612-1 Regulator 3 Self Open DM-V-274 Relief NNS Self Closed FW-FCV-510 Globe NNS Air Open FW-FCV-520 Globe NNS Air Open FW-FCV-530 Globe NNS Air Open FW-FCV-540 Globe NNS Air Open FW-LV-4210 Globe NNS Air Open FW-LV-4220 Globe NNS Air Open FW-LV-4230 Globe NNS Air Open FW-LV-4240 Globe NNS Air Open IA-V-545 Check 3 Delta P Open IA-V-546 Check 3 Delta P Open IA-V-547 Check 3 Delta P Open IA-V-548 Check 3 Delta P Open IA-V-549 Check 3 De lta P Open IA-V-550 Check 3 Delta P Open IA-V-8030 Check 3 Delta P Open IA-V-8031 Check 3 Delta P Open IA-V-8032 Check 3 Delta P Open IA-V-8033 Check 3 Delta P Open RC-LCV-459 Globe 1 Air Open RC-LCV-460 Globe 1 Air Open RMW-V-107 Relief NNS Self Closed SF-V-214 Check NNS Delta P Closed SF-V-215 Check NNS Delta P Closed SI-V-314 Relief NNS Self Closed SW-V-179 Check 3 Delta P Closed SW-V-180 Check 3 Delta P Closed SW-V-224 Check 3 Delta P Closed SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-28 Revision:

Sheet: 16 5 of 5 Tag No. Type ANS Safety Class Actuated By Normal Position SW-V-225 Check 3 Delta P Closed WLD-V-211 Reli ef NNS Self Closed WLD-V-277 Relief NNS Self Closed S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(B)-29 Revision:

Sheet: 9 1 of 1 TABLE 3.9(B)-29 S TRESS AND DEFLECTION ANALYSIS FOR THE SPENT FUEL POOL PUMPS (BINGHAM-WILLAMETTE CO

. 6X8X12 CF) Items - SSE Analysis Actual Allowable Natural Frequency (hz)

Pump 124 - Motor 190 - Pump Casing (psi) 2123 16500 Shaft (psi) 1317 28000 Pedestal Weld (psi)

Pump 3593 32400 Motor 93 32400 Hold Down Bolts (psi)

Pump 32035 70000 Motor 1179 70000 Base Hold Down Bolts (psi)

Tension 5160 19100 Shear 1930 9900 Clearances (in.)

Deflection @ coupling 0.0071 0.035 Deflection @ rotor 0.000662 0.0196*

  • Air gap between rotor/stator

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS T ABLE 3.9(N)-1 Revision:

Sheet: 10 1 of 2 Table 3.9(N)-1

SUMMARY

OF REACTOR COOLANT SYSTEM DESIGN TRANSIENTS Normal Conditions Occurrences

1. Heatup and cooldown at 100 F/hr (pressurizer cooldown 200 F/hr) 200 (each) 2. Unit loading and unloading at 5% of full power/min 13,200 (each) 3. Step load increase and decrease of 10% of full power 2,000 (each) 4. Large step load decrease with steam dump 200 5. Steady-state fluctuations : a. Initial fluctuations b. Random fluctuations

1.5x10 5 3.0x10 6 6. Feedwater cycling at hot shutdown 2000 7. Loop out of service: a. Normal loop shutdown

b. Normal loop startup

80 70 8. Feedwater Heaters Out of Service

a. One Heater Out of Service
b. One Bank of Heaters Out of Service 120 120 9. Unit loading and unloading between 0 to 15% of full power 500 (each)
10. Boron concentration equalization 26,400 11. Refueling 80
12. Reduced temperature return to power 2000 13. Reactor coolant pumps startup/shutdown 3800 14. Turbine roll test 20 15. Primary side leak test 200 16. Secondary side leak test 80 17. Tube leakage test 800 Upset Conditions 1. Loss of load, without immediate reactor trip 80 2. Loss of power (blackout with natural circulation in the Reactor Coolant System) 40 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS T ABLE 3.9(N)-1 Revision:

Sheet: 10 2 of 2 3. Partial loss of flow (loss of one pump) 80 4. Reactor trip from full power a. Without cooldown b. With cooldown, without safety injection

c. With cooldown and safety injection

230 160 10 5. Inadvertent reactor coolant depressurization 20 6. Inadvertent startup of an inactive loop 10 7. Control rod drop 80 8. Inadvertent emergency core cooling system actuation 60 9. Operating basic earthquake (5 earthquakes of 10 cycles each) 50 10. Excessive feedwater flow 30 11. RCS Cold Overpressurization 10 Emergency Condition* 1. Small loss-of-coolant accident 5 2. Small steam line break 5 3. Complete loss of flow 5 Faulted Conditions* 1. Main reactor coolant pipe break (large loss-of-coolant accident) 1 2. Large steam line break 1 3. Feedwater line break 1 4. Reactor coolant pump locked rotor 1 5. Control rod ejection 1 6. Steam generator tube rupture (included under upset conditions, reactor trip from full power with safety

injection) 7. Safe shutdown earthquake 1 Test Conditions 1. Primary side hydrostatic test 10 2. Secondary side hydrostatic test 10 *In accordance with the ASME Nuclear Power Plant Components Code, emergency and faulted conditions are not included in fatigue evaluation.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-2 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-2 LOADING COMBINATIONS FOR ASME CLASS 1 COMPONENTS A ND SUPPORTS Condition Classification Loading Combination Design Design Pressure, Design Temperature, Deadweight, Operating Basis Earthquake 1 Normal Normal Condition Transients, Deadweight Upset Upset Condition Transients, Deadweight, Operating Basis Earthquake Emergency Emergency Condition Transients, Deadweight Faulted Faulted Condition transients, Deadweight, Safe Shutdown Earthquake or (Safe Shutdown earthquake and Pipe Rupture Loads)

The operating basis earthquake is evaluated to ASME Level B stress limits for Class 1 piping components.

S EABROOK S TATION U F S A R D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS T ABLE 3.9(N)-3 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)-3 ALLOWABLE S TRESSES F OR A SME SECTION I II C LASS 1 C OMPONENTS (1) Operating Condition Classification Vessels/Tanks Piping Pumps Valves Component Supports Normal ASME Section III ASME Section III ASME Section III ASME Section III ASME Section III Subsection NF Upset ASME Section III ASME Section III ASME Section III ASME Section III ASME Section III Subsection NF Emergency ASME Section III ASME Section III ASME Section III ASME Section III ASME Section III Subsection NF Faulted ASME Section III See Subsection

3.9(N).1.4 ASME Section III See Subsection

3.9(N).1.4 ASME Section III See Subsection 3.9(N).1.4 (No active Class 1 pump used)

Note (2). ASME Section III Subsection NF See Subsection 3.9(N).1.4 P e , P m , P b , Q t , C p , S n & S m as defined by Section III, ASME Code Notes: (1) A test of the components may be performed in lieu of analysis.

(2) CLASS 1 VALVE FAULTED CONDITION CRITERIA ACTIVE a) Calculate Pm from para. NB3545.1 with Internal Pressure Ps = 1.25P Pm 1.5Sm b) Calculate Sn from para. NB3545.2 with Cp = 1.5 Ps = 1.25Ps Qt2 = 0 Ped = 1.3X value of Ped from equations of 3545.2(b)(1)

Sn 3Sm INACTIVE a) Calculate Pm from para. NB3545.1 with Internal Pressure Ps = 1.50 Ps Pm 2.4Sm or 0.7 Su b) Calculate Sn from para. NB3545.2 with

Cp = 1.5 Ps = 1.50Ps Qt2 = 0 Ped = 1.3X value of Ped from equations of 3545.2(b)(1)

Sn 3Sm SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-4 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-4 DESIGN LOADING COMBINATIONS FOR ASME CODE CLASS 2 A ND 3 COMPONENTS A ND SUPPORTS(1) Loading Combination(2,3) Design/Service Level Requirements

1. Design pressureDesign temperature, Dead weight Design 2. Normal condition pressure, normal condition metal temperature, deadweight, nozzle loads Service Level A
3. Upset condition pressure, Upset condition metal temperature, deadweight, nozzle loads, Operating Basis Earthquake Service Level B
4. Emergency condition pressure,emergency condition metal temperature, deadweight, nozzle loads Service Level C
5. Faulted condition pressure, faulted condition metal temperature, deadweight, nozzle loads, Safe Shutdown Earthquake Service Level D (1) Excludes active pumps. Refer to Table 3.9(N

)-8 for loading combinations and corresponding criteria applicable for active pumps.

(2) Temperature is used to determine allowable stress only.

(3) Nozzle loads, pressures, and temperatures are these associated with the respective plant operating conditions (i.e., normal, upset, emergency, and faulted) as noted, for the component under consideration.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-5 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-5 STRESS CRITERIA FOR SAFETY-RELATED ASME CLASS 2 (1) AND CLASS 3 TANKS Design/Service Level Stress Limits Design and Service Level A m 1.0 S m Lb 1.5 S Service Level B m 1.1 S m Lb 1.65 S Service Level C m 1.5 S m Lb 1.80 S Service Level D m 2.0 S m Lb 2.4 S (1) applies for tanks designed in accordance with ASME III, NC

-3300.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-6 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-6 STRESS CRITERIA FOR SAFETY-RELATED CLASS 2 TANKS (1) Design/Service Level Stress Limits Design and Service Level A P m 1.0 S m P L 1.5 S m (P m or P L) + P b 1.5 S m Service Level B P m 1.1 S m P L 1.65 S m (P m or P L) + P b 1.65 S m Service Level C P m 1.2 S m P L 1.8 S m (P m or P L) + P b 1.8 S m Service Level D P m 2.0 S m P L 3.0 S m (P m or P L) + P b 3.0 S m (1) Applies for tanks designed in accordance with ASME III, NC

-3200.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-7 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-7 STRESS CRITERIA FOR ASME CODE CLASS 2 A ND 3 INACTIVE PUMPS Design/Service Level Stress Limits Pmax* _____ Design and Service Level A m 1.0 S m Lb 1.5 S Service Level B m 1.1 S 1.1 m L) + b 1.65 S Service Level C m 1.5 S 1.2 m Lb 1.80 S Service Level D m 2.0 S 1.5 m Lb 2.4 S

  • The maximum pressure shall not exceed the tabulated factors listed under Pmax times the design pressure.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-8 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-8 DESIGN CRITERIA FOR ACTIVE PUMPS Design/Service Level Stress Limits Design, Service Level A and Service Level B m 1.0 S m b 1.5 S Service Level C m 1.2 S m b 1.65 S Service Level D m 1.2 S m + b 1.8 S SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-9 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-9 STRESS CRITERIA FOR SAFETY-RELATED ASME CODE CLASS 2 A ND CLASS 3 VALVES Design/Service Level Stress Limits (Notes 1

-4) Pmax (Note 5) Design & Service Valve bodies shall conform

1.0 Level

A to ASME Section III Service Level B m 1.1 S 1.1 m Lb 1.65 S Service Level C m 1.5 S 1.2 m Lb 1.80 S Service Level D m 2.0 S 1.5 m Lb 2.4 S Notes: 1. Valve nozzle (piping load) stress analysis is not required when both the following conditions are satisfied: (1) the section modulus and area of energy plane, normal to the flow, through the region defined as the valve body crotch are at least 110% of those for the piping connected (or join

-ed) to the valve body inlet and outlet nozzles; and, (2) code allowable stress, S, for valve body material is equal to or greater than the code allowable stress, S, for connected piping material. If the valve body material allowable stress is less than that of the connected

piping, the required acceptance criteria ratio shall be 110% multiplied by the ratio of the piping allowable stress to the valve allowable stress. If unable to comply with this requirement, an analysis in accordance with the design procedure for Class 1 valves is an acceptable alternate method.

2. Casting quality factor of 1.0 shall be used.
3. These stress limits are applicable to the pressure retaining boundary, and include the effects of loads transmitted by the extended structures, when applicable.
4. Design requirements listed in this table are not applicable to valve discs, stems, seat rings, or other parts of valves which are contained within the confines of the body and bonnet, or otherwise not part of the pressure boundary. 5. The maximum pressure resulting from upset, emergency or faulted conditions shall not exceed the tabulated factors listed under Pmax times the design pressure. If these pressure limits are met the stress limits in Table 3.9(N)-9 are considered to be satisfied.

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-10 Revision:

Sheet: 8 1 o f 1 TABLE 3.9(N)

-10 ACTIVE PUMPS Pump Item No. System ASME Safety Class Normal Mode Post LOCA Mode Function Centrifugal charging pump No. 1 CS-P-2A CVCS 2 ON/OFF ON High head safety injection Centrifugal Charging pump No. 2 CS-P-2B CVCS 2 ON/OFF ON High head safety injection Boric acid transfer pumps Nos. 1 and 2 CS-P-3A CS-P-3B BRS 2 ON/OFF OFF Boration and safe shutdown Residual heat removal pumps Nos. 1 and 2 RH-P-8A RH-P-8B RHRS 2 OFF ON Low head safety injection and normal cooldown Safety injec

-tion pumps Nos. 1 and 2 SI-P-6A SI-P-6B SIS 2 OFF ON Safety injection

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-11 Revision:

Sheet: 13 1 of 6 TABLE 3.9(N)-11 ACTIVE VALVES NSSS SUPPLIED Valve Number System Actuated by Size(in)

Type/ANS Safety Class RC-V-24 (1) RCS Self 3 Relief/2 RC-V-89 (1) RCS Self 3 Relief/2 RC-V-115 RCS Self Actuated 6 Safety/1 RC-V-116 RCS Self Actuated 6 Safety/1 RC-V-117 RCS Self Actuated 6 Safety/1 RC-V-122 RCS Motor 4 Gate/1 RC-V-124 RCS Motor 4 Gate/1 RC-PCV-456A RCS Solenoid 3x6 Relief/1 RC-PCV-456B RCS Solenoid 3x6 Relief/1 RMW-V-29 RCS P 3 Check/2 SF-V-197 (3) SF P 8 Check/3 CS-V-142 CVCS Motor 3 Gate/2 CS-V-143 CVCS Motor 3 Gate/2 CS-V-144 CVCS P 3 Check/2 CS-V-149 CVCS Motor 3 Gate/2 CS-V-150 CVCS Air 3 Globe/2 CS-V-167 CVCS Motor 2 Globe/2 CS-V-168 CVCS Motor 2 Globe/2 CS-V-173 CVCS Self 2 Relief/2 CS-V-178 CVCS P 3 Check/1 CS-V-179 CVCS P 3 Check/1 (1) This valve is required to perform its active safety function only during shutdown modes beyond hot standby. It is still included in the IST Program per the requirements of the ASME OM Code.

(3) Tested periodically by Sta tion procedures, but not included in the IST program because spent fuel pool cooling function does not meet ASME OM Code criteria.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-11 Revision:

Sheet: 13 2 of 6 Valve Number System Actuated by Size(in)

Type/ANS Safety Class CS-V-181 CVCS P 3 Check/1 CS-V-182 CVCS P 3 Check/1 CS-V-192 CVCS P 4 Check/2 CS-V-196 CVCS Motor 2 Globe/2 CS-V-197 CVCS Motor 2 Globe/2 CS-V-200 CVCS P 4 Check/2 CS-V-209 CVCS P 4 Check/2 CS-V-210 CVCS Manual 4 Gate/2 CS-V-219 CVCS Manual 3 Globe/2 CS-V-220 CVCS Manual 4 Gate/2 CS-V-221 CVCS Manual 3 Globe/2 CS-V-213 CVCS P 3 Check/2 CS-V-227 CVCS Self 3/4 Relief/2 CS-V-250 CVCS Self 2 Relief/2 CS-V-426 CVCS Motor 2 Globe/2 CS-V-437 CVCS Manual 4 Weir/3 CS-V-439 CVCS Manual 4 Weir/3 CS-V-440 CVCS P 4 Check/3 CS-V-1207 CVCS Manual 4 Weir/3 CBS-V-58 CVCS P 8 Check/2 CBS-V-60 CVCS P 8 Check/2 CS-LCV-112B CVCS Motor 4 Gate/2 CS-LCV-112C CVCS Motor 4 Gate/2 CS-LCV-112D CVCS Motor 8 Gate/2 CS-LCV-112E CVCS Motor 8 Gate/2 CO-V-434 EFW P 4 Check/3 CO-V-435 EFW P 4 Check/3 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-11 Revision:

Sheet: 13 3 of 6 Valve Number System Actuated by Size(in)

Type/ANS Safety Class IA-V-530 IA Air 2 Globe/2 RC-V-22 (2) RHRS Motor 12 Gate/1 RC-V-23 (2) RHRS Motor 12 Gate/1 RC-V-87 (2) RHRS Motor 12 Gate/1 RC-V-88 (2) RHRS Motor 12 Gate/1 RH-FCV-618 (2) RHRS Air 8 Butterfly/2 RH-FCV-619 (2) RHRS Air 8 Butterfly/2 RH-HCV-606 (2) RHRS Air 8 Butterfly/2 RH-HCV-607 (2) RHRS Air 8 Butterfly/2 RH-V-4 RHRS P 10 Check/2 RH-V-21 RHRS Motor 8 Gate/2 RH-V-22 RHRS Motor 8 Gate/2 RH-V-27 RHRS Air 3/4 Globe/2 RH-V-28 RHRS Air 3/4 Globe/2 RH-V-40 RHRS P 10 Check/2 RH-V-49 SIS Air 3/4 Globe/2 CBS-V-2 SIS Motor 12 Gate/2 CBS-V-5 SIS Motor 12 Gate/2 CBS-V-47 SIS Motor 8 Gate/2 CBS-V-48 SIS P 8 Check/2 CBS-V-49 SIS Motor 6 Gate/2 CBS-V-51 SIS Motor 8 Gate/2 CBS-V-52 SIS P 8 Check/2 CBS-V-53 SIS Motor 6 Gate/2 CBS-V-55 SIS P 12 Check/2 (2) This valve is repositioned during a plant cooldown beyond hot standby and is required to remain in its normal position during Modes 1, 2, and 3. The normal passive function must be included in the IST Program. Its active function may be periodically tested by another Station procedure.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-11 Revision:

Sheet: 13 4 of 6 Valve Number System Actuated by Size(in)

Type/ANS Safety Class CBS-V-62 SIS Self 3/4 Relief/2 CBS-V-145 SIS P 12 Check/2 CBS-V-56 SIS P 12 Check/2 CBS-V-146 SIS P 6 Check/2 CS-V-460 SIS Motor 6 Gate/2 CS-V-461 SIS Motor 6 Gate/2 CS-V-475 SIS Motor 6 Gate/2 NG-V-13 SIS Air 1 Globe/2 RH-V-13 SIS Self 3/4 Relief/2 RH-V-14 SIS Motor 8 Gate/2 RH-V-15 SIS P 6 Check/1 RH-V-25 SIS Self 3/4 Relief/2 RH-V-26 SIS Motor 8 Gate/2 RH-V-29 SIS P 6 Check/1 RH-V-30 SIS P 6 Check/1 RH-V-31 SIS P 6 Check/1 RH-V-32 SIS Motor 8 Gate/2 RH-V-35 SIS Motor 8 Gate/2 RH-V-36 SIS Motor 8 Gate/2 RH-V-50 SIS P 8 Check/1 RH-V-51 SIS P 8 Check/1 RH-V-52 SIS P 6 Check/1 RH-V-53 SIS P 6 Check/1 RH-V-70 SIS Motor 8 Gate/2 SI-V-3 (2) SIS Motor 10 Gate/1 SI-V-5 SIS P 10 Check/1 (2) This valve is repositioned during a plant cooldown beyond hot standby and is required to remain in its normal position during Modes 1, 2, and 3. The normal passive function must be included in the IST Program. Its active function may be periodically tested by another Station procedure.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-11 Revision:

Sheet: 13 5 of 6 Valve Number System Actuated by Size(in)

Type/ANS Safety Class SI-V-6 SIS P 10 Check/1 SI-V-10 SIS Self 1 Relief/2 SI-V-17 (2) SIS Motor 10 Gate/1 SI-V-20 SIS P 10 Check/1 SI-V-21 SIS P 10 Check/1 SI-V-30 SIS Self 1 Relief/2 SI-V-32 (2) SIS Motor 10 Gate/1 SI-V-35 SIS P 10 Check/1 SI-V-36 SIS P 12 Check/2 SI-V-45 SIS Self 1 Relief/2 SI-V-47 (2) SIS Motor 10 Gate/1 SI-V-50 SIS P 10 Check/1 SI-V-51 SIS P 10 Check/2 SI-V-60 SIS Self 1 Relief/2 SI-V-62 SIS Air 3/4 Globe/2 SI-V-70 SIS Air 3/4 Globe/2 SI-V-71 SIS P 4 Check/2 SI-V-76 SIS Self 3/4 Relief/2 SI-V-77 SIS Motor 4 Gate/2 SI-V-82 SIS P 6 Check/1 SI-V-87 SIS P 6 Check/1 SI-V-89 SIS Motor 11/2 Globe/2 SI-V-90 SIS Motor 11/2 Globe/2 SI-V-93 SIS Motor 2 Globe/2 SI-V-96 SIS P 4 Check/2 (2) This valve is repositioned during a plant cooldown beyond hot standby and is required to remain in its normal position during Modes 1, 2, and 3. The normal passive function must be included in the IST Program. Its active function may be periodically tested by another Station procedure.

S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-11 Revision:

Sheet: 13 6 of 6 Valve Number System Actuated by Size(in)

Type/ANS Safety Class SI-V-101 SIS Self 3/4 Relief/2 SI-V-102 SIS Motor 4 Gate/2 SI-V-111 SIS Motor 4 Gate/2 SI-V-112 SIS Motor 4 Gate/2 SI-V-113 SIS Self 3/4 Relief/2 SI-V-114 SIS Motor 4 Gate/2 SI-V-131 SIS Air 3/4 Globe/2 SI-V-134 SIS Air 3/4 Globe/2 SI-V-138 SIS Motor 4 Gate/2 SI-V-139 SIS Motor 4 Gate/2 SI-V-140 SIS P 3 Check/1 SI-V-157 SIS Air 3/4 Globe/2 SI-V-158 SIS Air 3/4 Globe/2 SI-V-160 SIS Air 3/4 Globe/2 SI-V-248 SIS Self 3/4 Relief/2

SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS EQUIPMENT AND SYSTEMS TABLE 3.9(N)-12 Revision:

Sheet: 8 1 of 1 TABLE 3.9(N)

-12 MAXIMUM DEFLECTIONS ALLOWED FOR REACTOR INTERNAL SUPPORT STRUCTURES Component Allowable Deflections (in)

No-Loss-of Function Deflections (in)

Upper Barrel Radial inward 4.1 8.2 Radial outward 1.0 1.0 Upper package 0.10 0.15 Rod Cluster guide tubes 1.00 1.75 S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.10 B-1 Revision:

Sheet: 14 1 of 3 TABLE 3.10B-1 SEISMIC QUALIFICATION METHOD

SUMMARY

OF INSTRUMENTATION AND ELECTRICAL EQUIPMENT CONTRACT NO. EQUIPMENT DESCRIPTION VENDOR QUALIFICATION METHOD REMARKS 106-2 Inverters - IE Elgar Analysis and Testing Wyle Lab 109-1 Cable Trays Metal Products Static Test and Analysis Metal Products 118-1 Penetrations Westinghouse Random Frequency Test and Analysis Acton Lab & W 118-3 Electrical Conductor Seal Assy. Conax Random Frequency Test Southwest Research Inst.

118-4 Electrical Connector Conax Random Frequency Test Southwest Research Inst.

119-3 Emergency Power Sequencer Brown Boveri Random Frequency Test Wyle Lab 119-5 DC Switchboards Brown Boveri Random Frequency Test Wyle Lab 120-1 Power Distribution Panels Gould Random Frequency Test Wyle Lab 120-9 R.C.P. Fuse Cabinets Powell Electric Random Frequency Test Southwest Research Inst. 120-11 Isolation Relay Cabinets Consolidated Controls Testing Farwell & Hendricks Inc.

129-1 Misc. Control Panels Systems Control Random Frequency Test 137-1 Batteries Gould Random Frequency Test and Analysis Wyle Lab and Gould 137-2 Battery Charges Power Conversion Products Random Frequency Test Wyle Lab 143-1 Motor Control Centers Brown Boveri Random Frequency Test Wyle Lab 144-1 5 kV and 15 Nonseg. Phase Bus Brown Boveri Analysis and Test Wyle Lab and Gould 145-2 5 kV Switchgear Brown Boveri Random Frequency Test Wyle Lab 145-3 480V Unit Substations Brown Boveri Random Frequency Test Wyle Lab S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.10 B-1 Revision:

Sheet: 14 2 of 3 CONTRACT NO. EQUIPMENT DESCRIPTION VENDOR QUALIFICATION METHOD REMARKS 162-1 Heat Tracing Therman Manufacturing Random Frequency Test Southwest Research Inst.

170-1 Main Control Board York Electro Panel Dynamic Analysis and Testing Analytical Eng. And Wyle Lab 170-4 Panel Mounted Small Case Recorders Foxbor o Random Frequency Test Southwest Research 170-5 Panel Mounted - Indicators Sigma International Instruments Random Frequency Test Acton Lab 170-6 Miscellaneous - I&C Panels Comsip Dynamic Analysis and Testing NQS, Analytical Eng. And Acton Lab 170-13 Intermediate Range - Nuclear Instr.

Class 1E Gamma-Metrics Random Frequency Test Wyle Lab 171-1 Instrument Racks Mercury Dynamic Analysis NQS 172-1 Radiation Data Management System General Atomic Random Frequency Test Wyle Lab 173-1 Control Valves Control Components Static Analysis 173-4 Solenoid Valves ASCO Size-bent Test 173-5 Control Valves - Nuclear Class 1, 2, 3 Masoneilan Static Analysis 173-7 Solenoid Valves Valcor Engineering Random Frequency Test Dayton T. Brown 174-1 Electronic Transmitters Foxboro Random Frequency Test Wyle Lab 174-2 Electronic Controllers & Accessories Westinghouse Random Frequency Test Westinghouse 174-8 MSIV Logic Cabinets Consolidated Controls Random Frequency Test Farwell & Hendricks 174-13 Electronic Transmitters Rosemount Random Frequency Test Wyle Lab S EABROOK S TATION UFSAR D ESIGN OF STRUCTURES , C OMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.10 B-1 Revision:

Sheet: 14 3 of 3 CONTRACT NO. EQUIPMENT DESCRIPTION VENDOR QUALIFICATION METHOD REMARKS 174-14 HELB Detection Isolation System Weed Instruments Analysis and Test 174-15 Level Transmitters Class 1E Transamerica-DeLaval Random Frequency Test Wyle Lab 252-3 Flow Nozzles and Tubes B.I.F. Static Analysis BIF 252-6 Rotometers Ametek/Shutte-Koerting Static analysis NUS 252-8 Thermal Elements & Wells Dravo/Thermo Electric Static Analysis Algor and Acton Lab 252-16 Differential Pressure Switches ITT Barton Random Frequency Test Approved Engineering Test Labs 252-19 Seismic Monitoring System Kinemetrics Random Frequency Test Wyle Lab 02209431 Loose Parts Monitoring System Physical Acoustic Corporation (PAC) /

MISTRAS Analysis and Test National Technical Systems 252-38 Temperature Switches(Class 1E) ASCO Sine Beat Test 501-1 Class 1E Hydrogen Analyzers Comsip Static Analysis and Test Engineering Analysis and Test SEABROOK STATION UFSAR DESIGN OF STRUCTURES , COMPONENTS E QUIPMENT AND SYSTEMS TABLE 3.10(N)-1 Revision:

Sheet: 8 1 of 1 TABLE 3.10(N)

-1 SEISMIC CATEGORY I INSTRUMENTATION A ND ELECTRICAL EQUIPMENT IN WESTINGHOUSE NSSS SCOPE O F SUPPLY Equipment EQDP* Safety-Related Valve Electric Motor Operators EQDP-HE-1 and 4 Safety-Related Pilot Solenoid Valves EQDP-HE-2 and 5 Safety-Related Externally Mounted Limit Switches EQDP-HE-3 and 6 Garrett (PORV) Solenoid

-Operated Pilot Valve EQDP-HE-9 Large Pump Motors (Outside Containment)

EQDP-AE-2 Canned Pump Motors (Outside Containment)

EQDP-AE-3 Electric Hydrogen Recombiner EQDP-SP-1 Pressure Transmitters EQDP-ESE-1 and 2 Differential Pressure Transmitters (Narrow Range)

EQDP-ESE-3 and 4 Resistance Temperature Detectors EQDP-ESE-5 and 6 Excore Neutron Detectors (Power Range)

EQDP-ESE-8 Nuclear Instrumentation System (NIS) Cabinet E QDP-ESE-10 Process Protection Sets EQDP-ESE-13 Solid-State Protection System and Safeguard Test Cabinet (2 Train)

EQDP-ESE-16 Instrument Bus Power Supply (Static Inverter)

EQDP-ESE-18 Reactor Trip Switchgear EQDP-ESE-20 Pressure Sensor

- Containment (Narrow Range)

EQDP-ESE-21 Pressure Sensor

- Containment (Wide Range)

EQDP-ESE-21 Differential Pressure Transmitters (Wide Range)

EQDP-ESE-4 *Equipment Qualification Data Package

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