ML18053A724
| ML18053A724 | |
| Person / Time | |
|---|---|
| Site: | Lee  |
| Issue date: | 12/19/2017 |
| From: | Donahue J Duke Energy Carolinas |
| To: | Office of New Reactors |
| Hughes B | |
| References | |
| DUKE, DUKE.SUBMISSION.15, LEE.NP, LEE.NP.1 | |
| Download: ML18053A724 (398) | |
Text
ports. These design criteria maintain structural integrity for seismic Category I and II ducts and tional capability for seismic Category I duct.
structural components of a typical HVAC duct system include the sheet metal ducts, stiffeners for ducts, duct supports, and other inline components such as duct heaters, dampers, etc.
1 Codes and Standards design of the HVAC ducts and their supports conform to the following codes and standards:
ASME N509-1989(R1996), Nuclear Power Plants Air Cleaning Units and Components ASME/ANSI AG-1-1997, Code on Nuclear Air and Gas Treatment American Institute of Steel Construction (AISC), Specification for the Design, Fabrication and Erection of Steel Safety Related Structures for Nuclear Facilities, AISC-N690-1994 American Iron and Steel Institute (AISI), Specification for the Design of Cold Formed Steel Structural Members, 1996 Edition and Supplement No. 1, July 30, 1999 SMACNA, HVAC Duct Construction Standards, Metal and Flexible, Second Edition 1995.
2 Loads and Load Combinations 2.1 Loads 2.1.1 Dead Load (D) d load includes the weight of the duct sheet, stiffeners and inline components such as duct ters and dampers. It also includes permanently attached items such as insulation and roofing, where applicable, and the weight of the duct supports. Temporary items used during struction or maintenance are removed prior to operation.
2.1.2 Construction Live Load (L) load consists of a load of 250 pounds to be applied only during construction or maintenance on rea of 10 square inches on the duct at a critical location to maximize flexural and shear stresses.
load is not combined with seismic loads.
2.1.3 Pressure (P) duct metal thickness and stiffener requirements are based on maximum system design sures. SMACNA or ASME guidelines, as applicable, are used in the design of duct metal kness and stiffener requirements.
pressure loads occur during normal plant operation, including plant start up testing, damper ure and normal airflow. Occasionally, overpressure transient loads such as rapid damper closure also produce short duration pressure differential.
3A-1 Revision 1
tation of the supports.
2.1.5 Wind Loads (W) twork within partially or fully vented buildings is subject to wind effects. Design wind loads are ussed in Section 3.3.
2.1.6 Tornado Loads (Wt) twork within partially or fully vented buildings is subject to tornado differential pressure effects.
ado loads are discussed in Section 3.3. Seismic Category I HVAC ductwork is protected from act by tornado missiles.
2.1.7 External Pressure Differential Loads (PA) mic Category I HVAC ductwork and its supports are designed to withstand dynamic external sure differential loads resulting from postulated accident conditions. Usually HVAC ducts are ed outside the areas of potential pipe break.
2.1.8 Thermal (TO/TA) sses on the supports resulting from the ductwork expansion due to temperature changes are ided by designing the system to take care of the expansion or by utilizing expansion joints. For ts of gasketed companion angle construction, thermal loads are negligible. For ducts exposed to er temperatures during a postulated accident condition, an evaluation is performed on a case by e basis for its effect.
2.2 Load Combinations load combinations for various service levels are as follows:
vice Level Load Combination (Construction / maintenance) D + L + P + TO (Normal Operating Condition) D + P + TO (Severe Condition) D + W + P + TO (Extreme Condition) D + Es + P + TO (Extreme Condition) D + Wt + P + TO (Abnormal Condition) D + P + PA + Es + TA 3 Analysis and Design HVAC duct support system is designed to maintain structural integrity of the duct. Function is not uired for the seismic Category II ductwork. The stresses are maintained within the allowable limits cified in Subsection 3A.3.4. Section properties and masses are calculated in accordance with ACNA standard.
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Bolted HVAC Ductwork 7 percent duct design due to pressure loads is based on ASME/ANSI AG-1 for seismic Category I ducts SMACNA for seismic Category II ducts.
global behavior of the duct is determined from the overall bending of the duct between the ports. It is similar to the beam type bending. The dead load is combined with the seismic inertial to determine the maximum bending moment. For determining the section modulus, the corners e duct are considered effective. The corner length in each direction equals 32 times the kness of the duct (t) for this purpose.
3.1 Response Due to Seismic Loads methodology for seismic analysis is provided in Subsection 3.7.3. Seismic loads are determined ither using the equivalent static load method of analysis or by performing dynamic analysis.
sses are determined for the seismic excitation in two horizontal and one vertical direction. The sses in the three directions are combined using the square root of sum of the squares (SRSS) hod or the 100-40-40 method as described in Subsection 3.7.3.6.
3.2 Deflection Criteria ections for panels and stiffeners conform to the limits stated in the Code on Nuclear Air and Gas atment.
3.3 Relative Movement arances are provided for allowing relative movement between equipment, other commodities, and C system.
3.4 Allowable Stresses basic stress allowables for the HVAC ducts are in accordance with paragraph SA-4220 of ME/ANSI AG-1.
basic stress allowables for duct supports utilizing rolled structural shapes are in accordance with SI/AISC N-690 and the supplemental requirements described in Subsection 3.8.4.5.2. The basic ss allowables for supports utilizing light gage cold rolled channel type sections are based on the ufacturer's published catalog values.
vice Level A and B Basic Allowable vice Level C and D 1.6 times basic allowable for tension and 1.4 times basic allowable for compression 3.5 Connections nections are designed in accordance with the applicable codes and standards listed in tion 3A.1. For connections used with light gage cold rolled channel type sections, design is based he manufacturer's published catalog values. Supports are attached to the building structure by ed or welded connections. Fastening of the supports to concrete structures meets the plemental requirements given in Subsection 3.8.4.5.1.
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igned to accommodate the effects of conditions associated with normal operation, anticipated sients, and postulated accident conditions. However, the dynamic effects associated with pipe ure may be excluded when analysis demonstrates that the probability of fluid system pipe rupture xtremely low. Dynamic effects are not considered for those segments of piping that are shown hanistically, with a large margin, not to be susceptible to a pipe rupture.
dynamic effects associated with pipe rupture include effects such as pipe break reaction loads, and jet impingement, subcompartment pressurization loads, and transient pipe rupture ressurization loads on other components.
use of mechanistic pipe break to eliminate evaluation of dynamic effects of pipe rupture includes erial selection, inspection, leak detection, and analysis. Subsection 3.6.3 outlines considerations tive to material selection, inspections, and leak detection. Subsection 5.2.5 describes the leak ction system inside containment. This appendix describes the analysis methods used to support application of mechanistic pipe break to high-energy piping in the AP1000.
analysis and criteria to eliminate dynamic effects of pipe breaks are encompassed in a hodology called leak-before-break (LBB). This methodology has been validated by theoretical stigations and test demonstrations sponsored by the industry and the NRC.
primary regulatory documents for leak-before-break analyses are General Design Criterion No. 4 C-4), Draft Standard Review Plan 3.6.3 (SRP 3.6.3) (Reference 1), and NUREG-1061, Volume 3 ference 2). Although SRP 3.6.3 has been issued only as a draft, its provisions are followed as elines to leak-before-break analyses.
k-before-break methodology has been applied to the reactor coolant loop and high-energy iliary line piping in operating nuclear power plants. The leak-before-break analysis used to port the piping design of the AP1000 is an application of the same methodology used in leak-re-beak evaluations previously accepted by the NRC.
e AP1000, leak-before-break evaluations are performed for the reactor coolant loop, the surge selected other branch lines containing reactor coolant down to and including 6-inch diameter inal pipe size, and portions of the main steam line. Those lines not qualified to the leak-before-ak criteria are evaluated using the pipe rupture protection criteria outlined in Subsections 3.6.1 3.6.2.
appendix provides a leak-before-break analysis for the applicable piping systems. Table 3B-1 ides a list of AP1000 leak-before-break piping systems.
1 Leak-before-Break Criteria for AP1000 Piping methodology used for leak-before-break analysis is consistent with that set forth in GDC-4, P 3.6.3 (Reference 1) and NUREG-1061, Volume 3 (Reference 2). The steps are:
Evaluate potential failure mechanisms Perform bounding analysis 3B-1 Revision 1
grity of the system as well as its suitability for leak-before-break analysis. The following lists ntial degradation (or "failure") mechanisms:
Erosion-corrosion induced wall thinning Stress corrosion cracking (SCC)
Water hammer Fatigue Thermal aging Thermal stratification Other mechanisms stainless steel piping is fabricated of SA312TP316LN or SA312TP304L material. The type 304L erial is used in the accumulator discharge lines. The main steam piping is fabricated of SA335 de P11. The welds are made by the gas tungsten arc welding (GTAW) method.
various degradation mechanisms are discussed in the following subsections.
2.1 Erosion-Corrosion Induced Wall Thinning mary Loop Piping l thinning by erosion and erosion-corrosion effects does not occur in the primary loop piping ause Series 300 austenitic stainless steel material is highly resistant to these effects. The coolant city in the AP1000 primary loop is about 76 feet per second. This flow velocity is not expected to te erosion-corrosion effects since stainless steels are considered to be virtually immune ference 3). A review of erosion-corrosion in nuclear power systems (Reference 4) reported that inless steels are increasingly being used due to their excellent resistance to erosion-corrosion, n at high water velocities, 40 m/s (131 ft/sec)." The bend radii in the AP1000 hot and cold legs are ater than the bend radii used in the crossover legs of operating plants. There is no record of ion-corrosion induced wall thinning in the primary loops of operating plants.
iliary Stainless Steel Piping l thinning by erosion-corrosion effects does not occur in the auxiliary stainless steel piping ause Series 300 austenitic stainless materials are highly resistant to these effects. The coolant city in these systems is lower than in comparable systems in operating Westinghouse-designed surized water reactors. There is no record of erosion-corrosion induced wall thinning in the nless steel piping of operating plants.
n Steam Line n steam lines in the AP1000 are fabricated from SA335 Grade P11 Alloy steel. Erosion-corrosion ced wall thinning is not expected in the main steam line. Extensive work has been done stigating erosion-corrosion in carbon steel pipes. The main steam line has low susceptibility to ion due to the pipe material composition, which has sufficient levels of chromium to preclude ion-corrosion material loss. Susceptibility is also low due to the relatively high operating perature and the high quality steam in the main steam line.
ed on the above discussion, erosion-corrosion induced wall thinning does not have an adverse ct on the integrity of the AP1000 leak-before-break piping systems.
3B-2 Revision 1
ditions necessary for stress corrosion cracking to take place are not present. If any of these three ditions is not present, stress corrosion cracking will not take place. The three conditions are:
There must be a corrosive environment.
The material itself must be susceptible.
Tensile stresses must be present in the material.
mary Loop Piping ing plant operation, the reactor coolant water chemistry is monitored and maintained within cific limits (see Subsection 5.2.3 for a discussion of reactor coolant chemistry). Contaminant centrations are kept below the thresholds known to be conducive to stress corrosion cracking.
major water chemistry control standards are included in the plant operating procedures as a dition for plant operation.
key to avoidance of a corrosive environment is control of oxygen. During normal power ration, oxygen concentration in the reactor coolant system is controlled to extremely low levels by trolling charging flow chemistry and maintaining a hydrogen overpressure in the reactor coolant at cified concentrations. Halogen concentration is controlled by maintaining concentrations of rides and fluorides within the specified limits. During plant operations, the likelihood of stress osion cracking in the primary loop piping systems is very low.
elements of a water environment known to increase the susceptibility of austenitic stainless steel tress corrosion are oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced s of sulfur (for example, sulfides, sulfites, and thionates). Pipe cleaning standards prior to ration and careful water chemistry control during plant operation are applied to prevent the urrence of a corrosive environment. Before being placed in service the piping is cleaned. During hes and preoperational testing, water chemistry is controlled according to written specifications.
ndards on chlorides, fluorides, conductivity, and pH are included in the guidelines for water for ning the piping.
es 300 stainless steel materials have been chosen for the AP1000 due to their proven operating erience. These materials have operated in low-oxygen or no-oxygen environments with no dents for a number of years. The requirements of Regulatory Guide 1.44 will be used to maintain experiences of the PWR applications for the use of Series 300 stainless steel materials.
ign tensile stresses in the reactor coolant loop are within the ASME Code,Section III allowables.
idual tensile stresses are expected in the welds and such stresses are not considered when igning by the ASME Code,Section III because these stresses are self-equilibrating and do not ct the failure loads. The residual stresses should not be more severe than for the operating stinghouse pressurized water reactor plants (which have not experienced stress corrosion king in the primary loop).
material used for buttering nozzles at the stainless-to-carbon steel safe ends is a high nickel
- y. The nickel-chromium-iron alloy selected and qualified for this application is not susceptible to ary water stress corrosion cracking.
iliary Stainless Steel Piping discussion above regarding the necessary conditions for primary loop piping stress corrosion king is also applicable to the other stainless steel piping of the primary system.
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experiences of the PWR applications for the use of Series 300 stainless steel materials.
ign tensile stresses in the other stainless steel piping are within the ASME Code,Section III wables. Residual tensile stresses are expected in the welds; however, the residual stresses uld not be more severe than for the operating Westinghouse pressurized water reactor plants ich have not experienced stress corrosion cracking in the auxiliary stainless steel piping).
n Steam Line main steam piping is constructed from ferritic steel. Stress corrosion cracking in ferritic steels monly result from a caustic environment. A source of a caustic environment in the main steam ng would be moisture carryover from the steam generator. However, the secondary side water tment utilizes all volatile treatment. All volatile treatment effectively precludes causticity in the m generator bulk liquid environment. For some operating plants prior to implementing all volatile tment, the phosphate water treatment caused a caustic chemical imbalance resulting in stress osion cracking of steam generator tubing. Under all volatile treatment water treatment conditions, e is no instance of caustic stress corrosion cracking on the ferritic steam lines indicating no ificant caustic carryover. The operating secondary side chemistry precludes stress corrosion king on the ferritic main steam line.
ed on the above discussion, stress corrosion cracking does not have an adverse effect on the grity of AP1000 leak-before-break piping systems.
2.3 Water Hammer mary Loop Piping reactor coolant loop is designed to operate at a pressure greater than the saturation pressure of coolant, thus precluding the voiding conditions necessary for water hammer to occur. The reactor lant primary system is designed for Level A, B, C, and D (normal, upset, emergency, and faulted) ice condition transients. The design requirements are conservative relative to both the number of sients and their severity. Relief valve actuation and the associated hydraulic transients following e opening have been considered in the system design. Other valve and pump actuations cause tively slow transients with no significant effect on the system dynamic loads.
rovide dynamic system stability, reactor coolant parameters are controlled. Temperature during mal operation is maintained within a narrow range by control rod positioning. Pressure is trolled within a narrow range for steady-state conditions by pressurizer heaters and pressurizer
- y. The flow characteristics of the system remain constant during a fuel cycle. The operating sients of the reactor coolant system primary loop piping are such that significant water hammer s are not expected to occur.
iliary Stainless Steel Piping passive core cooling system and automatic depressurization system are designed to minimize potential for water hammer induced dynamic loads. Design features include:
Continuously sloping core makeup tank and passive residual heat exchanger inlet lines to eliminate local high points Inlet diffusers in the core makeup tanks to preclude adverse steam and water interactions 3B-4 Revision 1
AP1000 pressurizer spray control valve is similar to what is used in the operating plants. There is istory of water hammer caused by the spray control valve.
normal residual heat removal system isolation valves are slow closing valves, identical to rating plants, and therefore would not be a source of water hammer.
se features minimize the potential of water hammer in the auxiliary stainless steel piping system.
n Steam Line steam lines are not subject to water hammer by the nature of the fluid transported. The following em design provisions address concerns regarding steam hammer within the main steam line and tify the significant dynamic loads included in the main steam piping design.
Design features that prevent water slug formations are included in the system design and layout. In the main steam system, these include the use of drain pots and the proper sloping of lines.
The operating and maintenance procedures that protect against a potential occurrence of steam hammer include system operating procedures that provide for slowly heating up (to avoid condensate formation from hotter steam on colder surfaces), operating procedures that caution against fast closing of the main steam isolation valves except when necessary, and operating and maintenance procedures that emphasize proper draining.
The stress analyses for the safety-related portion of the main steam system piping and components include the dynamic loads from rapid valve actuations, including actuation of the main steam isolation valves and the safety valves.
ed on the above discussion, water hammer does not have an adverse effect on the integrity of 000 leak-before-break piping systems.
2.4 Fatigue
-Cycle Fatigue
-cycle fatigue due to normal operation and anticipated transients is accounted for in the design of piping system. The Class 1 piping systems comply with the fatigue usage requirements of the ME Code,Section III. The Class 2 and 3 piping systems comply with the stress range reduction ors of the ASME Code,Section III.
to the nature of operating parameters, main steam line piping (Class 2) and the Class 3 portion e accumulator piping, are not subjected to any significant transients to cause low-cycle fatigue.
ed on the above discussion, low-cycle fatigue is not a concern of AP1000 leak-before-break ng systems.
h-Cycle Fatigue h-cycle fatigue loads in the system result primarily from pump vibrations. The steam generator is igned so that flow-induced vibrations in the tubes are avoided (see Subsection 5.4.2). The loads reactor coolant pump vibrations are minimized by criteria for pump shaft vibrations during hot tional testing and operation. During operation, an alarm signals when the reactor coolant pump ation is greater than the limits.
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2.5 Thermal Aging inless Steel Piping ng used in the reactor coolant loop and other auxiliary lines are wrought stainless steel materials, er than cast materials, so that thermal aging concerns are not expected for the AP1000 piping fittings. The welds used in the assembly of the AP1000 are gas tungsten arc welds (GTAW).
se welds are essentially as resistant to the effects of thermal aging as the base metal materials.
is due to the typically low ferrite contact in welds which results in minimal impact from thermal
- g. Based on this information, thermal aging of weld materials and piping used in the AP1000 is an issue.
n Steam Lines main steam piping system does not have cast materials. The welding process used on these s is also gas tungsten arc weld (GTAW).
re are no thermal aging concerns for the carbon steel piping of the main steam line and the alloy l of the main feedwater piping.
material used for the main steam piping system is not susceptible to dynamic strain aging cts.
2.6 Thermal Stratification k-before-break analyses include consideration of the loads and stresses due to thermal tification.
rmal stratification occurs only in a pipe that has a susceptible geometry and low flow velocities. A perature difference between the flowing fluid and stagnant fluid is also a prerequisite.
design of piping and component nozzles in the AP1000 includes provisions to minimize the ntial for and the effects of thermal stratification, cycling, and striping, pursuant to actions uested in several NRC bulletins, as discussed below.
mary Loop Piping rmal stratification in the reactor coolant loops resulting from actuation of passive safety features valuated as a design transient. Stratification effects due to both Level B and Level D service ditions are considered. The criteria used in the evaluation of the stress in the loop piping due to tification is the same as that applicable for other Level B and Level D service conditions.
iliary Stainless Steel Piping suant to the actions requested in NRC Bulletin 88-11, the pressurizer surge line is analyzed to onstrate that the applicable requirements of the ASME Code,Section III are met. This analysis udes consideration of plant operation, thermal stratification, and thermal striping using perature distributions and transients developed from experience on existing plant monitoring grams.
suant to the actions requested in NRC Bulletin 88-08 (cracking in piping connected to reactor lant systems due to isolation valve leakage), a systems review of the AP1000 piping was ormed in accordance with the criteria provided in Subsection 3.9.3.1.2.
3B-6 Revision 1
sive residual heat removal (PRHR) line from the hot leg, through the passive residual heat oval heat exchanger, and to the steam generator channel head potential for leakage through the isolation valves is not a concern for the piping extending from reactor coolant system hot leg connection to the passive residual heat removal heat exchanger
, since hot leakage from the reactor coolant system would be entering a hot section of piping.
kage exiting the passive residual heat removal heat exchanger would not be a concern since the led leakage would be entering a cold section of piping. This leakage would then heat up in the ng directly below the steam generator. Any amount of leakage is expected to be small, since the sure differential across the isolation valves is about 50 psi (the difference between the hot leg reactor coolant pump suction pressures). Activation of the passive residual heat removal system wing a plant scram is not a concern, since stratification will not occur due to the high flow velocity e passive residual heat removal return flow line.
omatic depressurization stage 4 lines from the hot legs to the stage 4 depressurization es kage is not a concern since the squib valves are leaktight and other potential leakage flow paths e double isolation.
ssurizer safety line from the pressurizer to the safety valve line is steam filled and will not experience stratified loadings.
omatic depressurization stage 2 and 3 lines from the pressurizer to the depressurization es kage is not a concern since double isolation exists in all potential leakage flow paths.
mal residual heat removal suction lines from the hot legs to the isolation valves rmal stratification in the normal residual heat removal suction lines, including leakage through the ation valves, is considered in the ASME pipe stress and fatigue analysis of these lines.
ct vessel injection lines rmal stratification in the direct vessel injection lines, including leakage through the isolation es, is considered in the ASME Code pipe stress and fatigue analysis of these lines.
n Steam Line steam lines are not subjected to thermal stratification by the nature of fluid transported.
ed on the above discussion, thermal stratification does not have an adverse effect on the integrity P1000 leak-before-break piping systems.
2.7 Other Mechanisms pipe evaluated for leak-before-break does not operate at temperature for which creep fatigue t be considered. Creep fatigue is a concern for ferritic steel piping operation at temperatures ve 700°F and for austenitic stainless steel operation above 800°F.
e degradation or failure by indirect causes such as fires, missiles, and component support failures recluded by criteria for design, fabrication, inspection, and separation of potential hazards in the nity of the safety-related piping. The structures, larger pipe, and components in the vicinity of pipe luated for leak-before-break are safety-related and seismically designed or are seismically ported if nonsafety-related.
3B-7 Revision 1
material fracture toughness tests.
3 Leak-before-Break Bounding Analysis methodology used for performing the bounding analysis is consistent with that set forth in C-4, SRP 3.6.3 (Reference 1) and NUREG-1061, Volume 3 (Reference 2).
nding leak-before-break analysis for the applicable AP1000 piping systems is performed. The lysis criteria and development techniques of the bounding analysis curves (BAC) are described
- w. The bounding analysis curve allows for the evaluation of the piping system in advance of the l piping analysis, incorporating leak-before-break considerations early in the piping design ess. The leak-before-break bounding analysis curve is used to evaluate critical points in the ng system. A minimum of two points are required to develop the bounding analysis curve. One t for the low normal stress case and the other point for the high normal stress case. If variations ipe size, material, pressure or temperature occur for a specific piping system, an additional nding analysis curve is generated. These points meet the following margins for leak-before-break lysis: (References 1 and 2).
Margin of 10 on leak detection capability Margin of 2 on flaw size Establish margin of 1 on load by using absolute combination method of maximum loads calculations to establish the bounding analysis curves use minimum values for wall thickness at weld counterbore and ASME Code material properties. For the main steam line lower bound erial property values determined from tests of the material are used. The use of the minimum es bounds the results of larger values. Since the piping is designed and analyzed using ASME e minimum material properties, these are used conservatively in a consistent manner for luation of leak-before-break evaluations. The as-built material properties are expected to be er than the ASME Code minimum properties. Using minimum thickness instead of a nominal kness is conservative for the stability analysis and was also used for leak-before-break in rating plants. The use of one thickness (either nominal or minimum) for both leak rate and stability ulation gives comparable overall margins for typical plant loads. The bounding analysis curves established using the axial load from internal pressure and neglecting other axial loads. This is an ropriate approximation because experience with leak-before-break calculations has shown that axial load due to pressure is the dominant axial load.
3.1 Procedure for Stainless Steel Piping 3.1.1 Pipe Geometry, Material and Operating Conditions following information is identified for each of the lines:
Piping materials - 316LN/304L, Type 304L is used for the accumulator discharge line Normal operating temperature Normal operating pressure Pipe outside diameter Pipe thickness 3B-8 Revision 1
- Pipe size
- Pipe schedule
- Operating pressures (100 percent power and maximum stress condition)
- Operating temperatures (100 percent power and maximum stress condition) 3.1.2 Pipe Physical Properties physical and metallurgical properties for each of the lines are determined in the following manner Minimum wall thickness is calculated at the weld counterbore The area (A) and section modulus (Z) are calculated using minimum wall thickness The yield strength is the ASME Code,Section II (Reference 5) minimum value, at temperature of interest The ultimate strength is the ASME Code,Section II (Reference 5) minimum value, at temperature of interest The modulus of elasticity is the ASME Code,Section II (Reference 5) at temperature of interest 3.1.3 Low Normal Stress Case (Case 1) etermine the first point of the bounding analysis curve the following steps are used.
Calculate axial force Fp (for normal operating pressure)
Assume a lower magnitude of bending stress. The magnitude selected is a very small number that is lower than the expected minimum bending stress.
Calculate bending moment = (bending stress) x (section modulus)
Calculate the leakage flaw size at 100 percent power condition for 10 times the leak detection capability (for 0.5 gpm leak detection capability, this is 10 x 0.5 = 5 gpm)
Perform the stability analysis using the limit load methodology to obtain the critical flaw size.
For AP1000 piping systems, there is no cast material and the weld process is gas tungsten arc welds (Z factor is 1.0 since weld process is gas tungsten arc welds, Reference 1.)
- Determine the maximum loads for a critical flaw size of twice the leakage flaw size. The margin of 2 on flaw size is satisfied.
Calculate the low normal stress and corresponding maximum stress by using:
Axial Force Bending Moment Stress = + (3B-1)
Area Section Modulus 3B-9 Revision 1
Axial force Fp is calculated as above for normal operating pressure Assume a higher magnitude of bending stress to get higher bending moment. The magnitude of bending is selected such that the corresponding maximum stress generated is close to the flow stress.
Calculate bending moment = (bending stress) x (section modulus)
Repeat leakage flaw size and stability calculations as outlined for the low normal stress case above e: For an intermediate point, calculation steps are the same as low normal or the high normal case.
3.1.5 Develop the Bounding Analysis Curve For Case 1, normal and maximum stresses are established.
For Case 2, normal and maximum stresses are established.
Plot these two points with normal versus maximum stress. The curve is generated by joining these two points in a straight line. More than two points may be used if desired, to obtain a smooth curve fit between the calculated points. A typical curve is shown in Figure 3B-1.
3.2 Procedure for Non-stainless Steel Piping procedure to develop the bounding analysis curve for the carbon steel for main steam lines is lar to that for the stainless steel and is described below.
3.2.1 Pipe Geometry, Material and Operating Conditions following information is identified for each of the lines:
Piping materials Normal operating temperature Normal operating pressure Pipe outside diameter Piping thickness number of bounding analysis curves needed for each analyzable piping system is determined by view of the combinations of the following parameters:
- Pipe size
- Pipe schedule
- Operating pressures (100 percent power and maximum stress condition)
- Operating temperatures (100 percent power and maximum stress condition) 3B-10 Revision 1
The area (A) and section modulus (Z) are calculated using minimum wall thickness The material yield strength, ultimate strength, modulus of elasticity, stress-strain curves, and J-R curves are determined from the material tests 3.2.3 Low Normal Stress Case (Case 1) etermine the first point of the bounding analysis curve the following steps are used.
Calculate axial force Fp (for normal operating pressure)
Assume a lower magnitude of bending stress Calculate bending moment = (bending stress) x (section modulus)
Calculate the leakage flaw size at 100 percent power condition for 10 times the leak detection capability (for 0.5 gpm leak detection capability, this is 10 x 0.5 = 5 gpm)
Stability analysis
- Perform J-integral analysis
- Determine the maximum loads for a critical flaw size of twice the leakage flaw size by satisfying the stability criteria. The margin of 2 on flaw size is satisfied.
Stability criteria
- Japplied JIC
- If Japplied > JIC, then Japplied < Jmax and Tapplied < Tmat Calculate the low normal stress and corresponding maximum stress by using:
Axial Force Bending Moment Stress = +
Area Section Modulus 3.2.4 High Normal Stress Case (Case 2) etermine the other endpoint of the bounding analysis curve the following steps are used.
Axial force Fp is calculated above (for normal operating pressure)
Assume a higher magnitude of bending stress to get higher bending moment Calculate bending moment = (bending stress) x (section modulus) 3B-11 Revision 1
e: For an intermediate point, calculation steps are the same as low normal or the high normal case.
3.2.5 Develop the Bounding Analysis Curve ow steps as outlined for the stainless steel case in Subsection 3B.3.1.5.
3.3 Evaluation of Piping System Using Bounding Analysis Curves valuate the applicability of leak-before-break, the results of the pipe stress analysis are pared to the bounding analysis curve. The critical location is the location of highest maximum ss as determined by the pipe stress results. A comparison is made with the applicable bounding lysis curves for the analyzable piping systems. As outlined in 3B.3.1.1 and 3B.3.2.1, bounding lysis curves are calculated for different combinations of pipe size, pipe schedule, operating sures, operating temperatures.
bounding analysis curves are used during the layout and design of the piping systems to provide sign that satisfies leak-before-break criteria. In addition, the results of the as-built piping analysis nciliation to the bounding analysis curves to verify that the fabricated piping systems satisfy leak-re-break criteria. See Subsection 3.6.4 for the Combined License information item associated this verification.
he critical location, the load combination for the maximum stress calculation uses the absolute method. The load combination is as follows:
Pressure + Deadweight + Thermal (100% Power)* + Safe Shutdown Earthquake normal stress is calculated using the algebraic sum method at critical location and the following combination.
Pressure + Deadweight + Thermal (100% Power*)
cludes applicable stratification loads.
3.3.1 Calculation of Stresses stresses due to axial loads and moments are calculated by the following equation:
re:
F M
= + (3B-2)
A Z
= stress F = axial load M = moment A = cross-sectional area Z = section modulus 3B-12 Revision 1
re, M = moment for required loading MX = torsional moment MY = Y component of bending moment MZ = Z component of bending moment Y and Z-axes are lateral axes to the X-axis which is the axial axis axial load and moments for the normal case and maximum case are computed by the methods wn below.
3.3.2 Normal Loads normal operating loads are calculated by the following equations:
MZ = (MZ)DW + (MZ)Th (3B-7) subscripts of the above equations represent the following load cases:
DW = deadweight Th = normal thermal expansion (100 percent power, including applicable stratification loads)
P = load due to internal pressure method of combining loads is often referred to as the algebraic sum method.
culate the normal stress at the critical location.
3.3.3 Maximum Loads the maximum case, the absolute summation method of load combination is applied which results igher magnitude of the combined loads. Since stability is demonstrated using these loads, the
-before-break margin on loads is satisfied. An example of the absolute summation expressions shown below:
F = FDW + FTh + FP + FSSEINERTIA + FSSEAM (3B-8)
MX = (MX)DW + (MX)Th + (MX)SSEINERTIA + (MX)SSEAM (3B-9) 3B-13 Revision 1
re subscripts SSE, Inertia and AM mean safe shutdown earthquake, inertia and anchor motion ectively.
3.3.4 Bounding Analysis Curve Comparison - LBB Criteria ompare the stress results with the bounding analysis curve the following process is followed. The mal and maximum stress at the critical location are calculated by using the loads defined in section 3B.3.3. Plot the normal stress versus maximum stress on the bounding analysis curve for specified system. If the point is on or below the bounding analysis curve, the leak-before-break lysis and margins are satisfied. If the point falls above the bounding analysis curve, the leak-re-break analysis criteria are not satisfied and the pipe layout or support configuration needs to evised to meet the leak-before-break bounding analysis. Figure 3B-1 shows a typical bounding lysis curve.
3.4 Bounding Analysis Results le 3B-1 shows a summary of piping systems and corresponding bounding analysis figures.
res 3B-1 to 3B-22 show the bounding analysis curves. The curves satisfy the margins as cated in Section 3B.3.
4 Differences in Leak-before-Break Analysis for Stainless Steel and Ferritic Steel Pipe significant difference between leak-before-break analysis performed for the stainless steel ss 1 and Class 3) systems and the ferritic steel in the Class 2 systems is in the stability analysis.
e case of stainless steel systems, stability analyses are performed by limit load approach. In the tic steel systems, stability analyses are performed by J-integral approach.
5 Differences in Inspection Criteria for Class 1, 2, and 3 Systems ss 1, 2 and 3 systems are subjected to in-service inspection requirements from ASME Code, tion XI. For Class 1 piping, terminal ends and dissimilar metal welds are volumetrically inspected, g with other locations, to total 25 percent of the welds. For Class 2 piping, the requirement is to metrically inspect the terminal ends and other locations to total 7.5 percent of the welds. For ss 3 systems (the only Class 3 piping is in the accumulator line which is always at room perature), the system receives periodic visual examinations in conjunction with pressure testing.
se requirements were developed by ASME Code,Section XI consistent with the different safety ses of these systems.
leak-before-break evaluations are based on the ability to detect a potential leaking crack; not the ty to find cracks by inservice inspections. The criteria or methods of the leak-before-break luations are the same for ASME Code Class 1, 2, and 3.
6 Differences in Fabrication Requirements of ASME Class 1, Class 2, and Class 3 Piping significant difference among Class 1, 2 and 3 seamless pipe occurs in the nondestructive mination requirements. The Class 1 seamless pipe examination requirements include an sonic testing examination, whereas Class 2 and 3 do not. In addition, the Class 1 examination uirements for a circumferential butt welded joint include radioagraphic testing and magnetic 3B-14 Revision 1
ography will be conducted on a random sample of welds. The Class 3 leak-before-break lines are uded in the lines that are radiographed. In addition see Subsection 3.6.3.2 for augmented ection of Class 3 leak-before-break lines.
the fabrication of welds in the Class 1, Class 2 and Class 3 pipes there is no significant rences.
differences in fabrication and nondestructive examination requirements do not affect the leak-re-break analyses assumptions, criteria, or methods.
7 Sensitivity Study for the Constraint Effect on LBB stinghouse performed a sensitivity study on a 6-inch diameter pipe to demonstrate that the leak-re-break evaluation margins are not significantly affected when constraint effects of pressure ced bending are included. The analysis used a finite element model of a 6-inch diameter pipe ded to a nozzle with a fixed end condition. This conservatively represents the bounding conditions AP1000 piping. The normal and maximum stresses were used from a representative AP600 ch line bounding analysis curve. The material properties for the base metal and TIG weld were sidered in the analysis. The stability analysis was performed using the J-integral method. This lysis was developed in consultation with the NRC.
conclusion of this sensitivity study is that the leak-before-break margins for 6-inch and larger ng on AP1000 are not significantly affected by the constraint effect and application of leak-before-ak to such piping is acceptable.
8 References Standard Review Plan 3.6.3, "Leak Before Break Evaluation Procedures," Federal Register, Volume 52, Number 167, Friday, August 28, 1987; Notice (Public Comment Solicited), pp. 32626-32633.
NUREG-1061, "Evaluation of Potential for Pipe Breaks, Report of the U.S. Nuclear Regulatory Commission Piping Review Committee," Volume 3, (prepared by the Pipe Break Task Group), November 1984.
"Erosion-Corrosion in Nuclear Plant Steam Piping: Causes and Inspection Program Guidelines," EPRI NP-3944, April 1985.
G. Cragnolino, "Erosion-Corrosion in Nuclear Power Systems-An Overview,"
Corrosion '87, Paper No. 86, March 1987.
ASME Boiler and Pressure Vessel Code,Section II, "Materials," 1998 Edition through 2000 Addenda.
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Nominal Diameter Temp Pressure tem Subsystem Line No(s). (Inches) Material (°F) (psig) Figure No.
(1) SA-376 TP316LN 610.0 2248 3B-2 CS Primary Loop Hot Leg L001A, B 31 (ID)
CS Primary Loop Cold Leg L002A, B, C, D 22 (ID)(1) SA-376 TP316LN 537.2 2310 3B-3 GS Main Steam Line L006A, B 38 SA-335 GR P11 523.0 821 3B-4 CS Normal Residual Heat Removal L139 20 SA-312 TP316LN 610.0 2248 3B-5 CS Surge Line L003 18 SA-312 TP316LN 653.0 2248 3B-6 (Sheet 1)
CS Surge Line L003 18 SA-312 TP316LN 455.0 430 3B-6 (Sheet 2)
CS Passive Residual Heat Removal Supply/ L135A,B; L136A,B 18 SA-312 TP316LN 610.0 2248 3B-7 ADS 4 CS Passive Removal Heat Removal Supply/ L133A, B; L137A, B; L134 14 SA-312 TP316LN 610.0 2248 3B-8 ADS 4 XS Passive Residual Heat Removal Supply to L102, L107 14 SA-312 TP316LN 610.0 2248 3B-8 Cold Trap and Vent Line XS Passive Residual Heat Removal Supply L102 14 SA-312 TP316LN 120.0 2248 3B-9 after Cold Trap to PRHR HX XS Return - PRHR HX to Isolation Valve L103; L104A, B 14 SA-312 TP316LN 120.0 2248 3B-9 CS Automatic Depressurization System L004A,B; L006A,B; L020A,B; 14 SA-312 TP316LN 653.0 2235 3B-10 Stage 2, 3 L030A, B; L131 XS Passive Residual Heat Removal Return - L104A, B; L105 14 SA-312 TP316LN 537.0 2190 3B-11 after Isolation Valve 3B-16 Revision 1
Nominal Diameter Temp Pressure tem Subsystem Line No(s). (Inches) Material (°F) (psig) Figure No.
CS Passive Residual Heat Removal Return L113 14 SA-312 TP316LN 537.0 2190 3B-11 XS Passive Residual Heat Removal Vent Line L107 12 SA-312 TP316LN 610.0 2248 3B-12 (Not Used)
XS Accumulator to Isolation Valve L029A, B 8 SA-312 TP304L 120.0 700 3B-13 CS Balance Line from Cold Leg to CMT L118A, B 8 SA-312 TP316LN 537.0 2310 3B-14 Isolation Valve XS Balance Line from CMT Isolation Valve to L007A, B; L070A, B 8 SA-312 TP316LN 537.0 2310 3B-14 CMT XS Direct Vessel Injection Line to RV L021A, B; L125A, B 8 SA-312 TP316LN 537.0 2310 3B-14 XS Core Makeup Tank (Injection Line, RV Side L015, L016, L017, L018, 8 SA-312 TP316LN 120.0 2310 3B-15 of Isolation Valve, Core Makeup Tank Side L020, L021, L025, L127 of Isolation Valve), Direct Vessel Injection (Accumulator Connection to Cold Trap), (All A, B)
IWRST Injection CS Automatic Depressurization System L021A,B; L031A,B 8 SA-312 TP316LN 653.0 2235 3B-16 (Not Stage 2, 3 Used)
XS Accumulator after Isolation Valve L027A, B 8 SA-312 TP304L 120.0 700 3B-17 XS RNS Discharge L019A, B 6 SA-312 TP316LN 120.0 2310 3B-18 CS Automatic Depressurization System Header L005A, B 6 SA-312 TP316LN 653.0 2235 3B-19 to RCS Safety Valve CS Normal Residual Heat Removal L140 12 SA-312 TP316LN 610.0 2248 3B-20 NS Normal Residual Heat Removal L001, L002A, B 10 SA-312 TP316LN 610.0 2248 3B-21 CS Automatic Depressurization System L021A, B; L031A, B 8 SA-312TP316LN 250 2235 3B-22 Stage 2, 3 (Cold Trap)
ID = Inside diameter 3B-17 Revision 1
Figure 3B-1 Typical Bounding Analysis Curve (BAC) 3B-18 Revision 1
Figure 3B-2 Bounding Analysis Curve for Primary Loop Hot Leg 3B-19 Revision 1
Figure 3B-3 Bounding Analysis Curve for Primary Loop Cold Leg 3B-20 Revision 1
Figure 3B-4 Bounding Analysis Curve for 38 Main Steam Line 3B-21 Revision 1
Figure 3B-5 Bounding Analysis Curve for 20 Normal RHR 3B-22 Revision 1
Figure 3B-6 (Sheet 1 of 2)
Bounding Analysis Curve for 18 Surge Line 3B-23 Revision 1
Figure 3B-6 (Sheet 2 of 2)
Bounding Analysis Curve for 18 Surge Line 3B-24 Revision 1
Figure 3B-7 Bounding Analysis Curve for 18 PRHR Supply/ADS 4 3B-25 Revision 1
Figure 3B-8 Bounding Analysis Curve for 14 PRHR Supply to Cold Trap, PRHR Supply/ADS4 3B-26 Revision 1
Figure 3B-9 unding Analysis Curve for 14 PRHR Supply after Cold Trap, Return - to Isolation Valve 3B-27 Revision 1
Figure 3B-10 Bounding Analysis Curve for 14 ADS Stage 2, 3 3B-28 Revision 1
Figure 3B-11 ounding Analysis Curve for 14 PRHR Return - after Isolation Valve, 14 PRHR Return 3B-29 Revision 1
3B-30 Revision 1 Figure 3B-13 Bounding Analysis Curve for 8 Accumulator to Isolation Valve 3B-31 Revision 1
Figure 3B-14 unding Analysis Curve for 8 CMT Cold Leg Balance Line and Vent, DVI Cold Trap to RV 3B-32 Revision 1
Figure 3B-15 Bounding Analysis Curve for 8 CMT, DVI IWRST (Various Sections) 3B-33 Revision 1
3B-34 Revision 1 Figure 3B-17 Bounding Analysis Curve for Accumulator after Isolation Valve 3B-35 Revision 1
Figure 3B-18 Bounding Analysis Curve for RNS Discharge 3B-36 Revision 1
Figure 3B-19 Bounding Analysis Curve for ADS Header to RCS Safety Valve 3B-37 Revision 1
Figure 3B-20 Bounding Analysis Curve for 12 Normal RHR 3B-38 Revision 1
Figure 3B-21 Bounding Analysis Curve for 10 Normal RHR 3B-39 Revision 1
Figure 3B-22 Bounding Analysis Curve for 8 ADS Stage 2, 3 3B-40 Revision 1
s, beams, elbows, masses, and springs. The structural model is subjected to internal pressure, mal expansion, weight, seismic, and pipe break loadings with imposed boundary conditions. The e element displacement method is used for the analysis. The stiffness matrix for each element is embled into a system of simultaneous linear equations for the entire structure. This set of ations is then solved by a variation of the Gaussian elimination method, known as the wave-front nique. This technique makes it possible to solve systems of equations with a large number of rees of freedom using a minimum amount of computer memory.
1 Reactor Coolant Loop Model Description piping model of the reactor coolant loop consists of a number of elements of given dimensions, s, and physical properties that mathematically simulate the structural response of the physical em. The system model contains the reactor pressure vessel (RPV), two steam generators (SGs),
reactor coolant pumps (RCPs), the reactor coolant loop piping, and the primary equipment ports. A two-loop model is developed for the AP1000 reactor coolant loop system.
stiffness and mass effects of branch piping connected to the primary loop piping are considered n significant (Subsection 3.7.3.8.1).
1.1 Steam Generator Model 1.1.1 Steam Generator Mass and Geometrical Model steam generator is represented by discrete masses. The geometry of the steam generator sel is used to determine the properties of the equivalent piping elements that join the steam erator masses for sections of the steam generator above the tubesheet. For the steam generator nnel head, a super element is used to represent the stiffness characteristics that link the steam erator lower shell with the steam generator supports and nozzles. The modulus of elasticity and fficient of thermal expansion corresponding to the thermal conditions are applied to the steam erator equivalent piping elements.
1.1.2 Steam Generator Supports values of the steam generator support stiffnesses and locations of the supports are determined the finite element models of the support members. The stiffness of the upper lateral supports ude the steam generator shell flexibility. The local concrete building flexibility is included in the port stiffness.
1.2 Reactor Coolant Pump Model 1.2.1 Static Model reactor coolant pump is represented by a super element to represent the mass and stiffness racteristics of the pump. For a thermal expansion analysis, rigid links are modeled in parallel with per element with the thermal expansion coefficient incorporated.
1.2.2 Seismic Model reactor coolant pump is represented by a super element to represent the mass and stiffness racteristics of the pump. The reactor coolant pump model is a detailed model similar to that used ualify the pump.
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erator channel head in each of the reactor coolant loops.
1.3 Reactor Pressure Vessel Model 1.3.1 Mass and Geometrical Model reactor pressure vessel model consists of equivalent pipe, stiffness, and mass elements. The ments represent the vessel shell, the vessel core barrel, the fuel assemblies, and the integrated d lift package.
reactor pressure vessel is modeled with equivalent pipe elements and connecting stiffnesses.
equivalent pipe element properties of the vessel and barrel are those of the cylindrical structures.
beam properties of the reactor internals are adjusted to simulate their fundamental frequency.
appropriate modulus of elasticity and coefficient of thermal expansion are used for the equivalent elements representing the reactor pressure vessel.
1.3.2 Reactor Pressure Vessel Supports reactor pressure vessel is supported at the four reactor pressure vessel inlet nozzles. Each port consists of a vertical stiffness and a lateral tangential stiffness. The support is represented by iffness matrix. The reactor pressure vessel supports are active for the analyzed loading ditions. The reactor pressure vessel model includes the effects of the vessel shell flexibility at the and outlet nozzles. The local concrete building flexibility is included in the support stiffness.
1.4 Containment Interior Building Structure Model ntainment interior building structure finite element model is not required because the seismic ts to the reactor coolant loop model are provided at all of the building attachments to the reactor lant loop.
1.5 Reactor Coolant Loop Piping Model reactor coolant loop piping model consists of piping elements and bends. Each reactor coolant has two cold legs and one hot leg. The straight runs and bends of the cold leg and hot leg are t with the nominal dimensions. Each reactor coolant loop branch connection is represented by a e point. The reactor coolant loop piping model contains distributed masses of the hot and cold leg ng for static deadweight analysis and lumped masses representing the hot and cold leg piping for amic analysis.
2 Design Requirements reactor coolant piping is qualified to the requirements of the ASME Code,Section III, section NB, 1989 Edition with 1989 Addenda.
loadings for ASME Code,Section III, Class 1 components are defined in Subsection 3.9.3. The wing loadings are considered in the reactor coolant loop piping analysis:
Design pressure (P)
Safe shutdown earthquake (SSE)
Design basis pipe break (DBPB)
Building motions due to automatic depressurization system sparger discharge into the IRWST Thermal stratification during transient conditions ddition to the analyses of these loads, the reactor coolant piping is analyzed for the effect of cyclic ue due to the design transients and earthquakes smaller than SSE.
3 Static Analyses 3.1 Deadweight Analysis reactor coolant loop piping system is analyzed for the effect of deadweight. The deadweight lysis is performed without considering the dry weight of the directly supported equipment. The cts of the auxiliary branch piping on the reactor coolant loop are generally negligible by the design e auxiliary supports. A deadweight analysis is performed to include the total weight of the reactor lant loop piping and the water weight in the components.
reactor coolant loop deadweight model includes the corresponding active reactor coolant loop ports - reactor pressure vessel supports, and the steam generator column and lower and rmediate lateral strut supports. The steam generator upper lateral snubber supports are sidered as inactive.
3.2 Internal Pressure Analysis effects of the internal primary coolant pipe pressure are used in the calculations of forces and ments for both the reactor coolant loop piping and equipment supports. The moment stress due to sure is considered negligible for the ASME Code pipe stress equations.
3.3 Thermal Expansion Analysis reactor coolant loop piping is analyzed for the effects of thermal expansion. The thermal ansion analysis model considers the expansion of the reactor coolant loop piping, reactor sure vessel, steam generator, reactor coolant pump, and the equipment supports. The stiffness cts of the auxiliary piping on the reactor coolant loop expansion are generally negligible by the ign of the auxiliary lines supports.
4 Seismic Analyses reactor coolant loop piping is analyzed for the dynamic effects of a safe shutdown earthquake E).
model used in the static analysis is modified for the dynamic analysis by including the lumped s characteristics of the piping and equipment. The effect of the equipment motion on the reactor lant loop piping and support system is obtained by modeling the mass and stiffness racteristics of the equipment in the overall system model. The reactor coolant loop seismic 3C-3 Revision 1
time history integration method of analysis is used for the reactor coolant loops. The seismic t considers the soil profiles described in Subsection 3.7.1. This input is obtained from the nuclear nd seismic analysis with time history input generated from the enveloped basemat response ctra of the soil cases described in Subsection 3.7.1. The duration of the input is between 12 to econds, depending on the duration needed to envelop the design response spectra. Three runs e performed based on the envelope of the soil profiles, the building model at nominal stiffness, at stiffness varied by + or - 30 percent to account for uncertainties. The reactor coolant loop uses arate time history displacement input from the building analysis at the primary support locations.
direct integration is used with Rayleigh damping for loop components at 4 percent of critical ping. The steam generator snubbers have different stiffnesses in tension and compression. The n value of the tension and compression stiffness is used in order to keep the model linear. The tor pressure vessel vertical supports are acting downward only and are preloaded by dweight, pressure, and thermal expansion loadings. The time history analysis is performed to luate the effect of lift-off of the vessel at the location of these supports.
5 Reactor Coolant Loop Piping Stresses revent gross rupture of the reactor coolant loop piping system, the general and local primary mbrane stress criteria must be satisfied. This is accomplished by satisfying Equation (9) in agraph NB-3652 of the ASME Code,Section III. The secondary stress caused by thermal ansion is qualified by satisfying Equation (12) in paragraph NB-3653 of the ASME Code, tion III.
6 Description of Computer Programs section provides a list of computer codes used for the AP1000 reactor coolant loop system lysis. Brief descriptions of the functions of each computer code are the following:
SYS - Performs Structural Analysis Using Finite Element Analysis Method. Displacements and s are calculated at the pipe elements, supports, and equipment nozzles for pressure, dweight, thermal, and seismic loadings.
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ety-related electrical equipment is tested under the environmental conditions expected to occur in event of a design basis event. This testing provides a high degree of confidence in the safety-ted system performance under the limiting environmental conditions. Qualification criteria were sed by IEEE 323-1974 (Reference 1) and by Regulatory Guide 1.89, which endorses this IEEE dard. The concept of aging was highlighted in IEEE 323-1974, and interpretation of the scope of g and implementation methods were subsequently developed. 10 CFR 50.49 provides the NRC uirements for qualification of equipment located in potentially harsh environments. Therefore, the ance provided by IEEE 323-1974 is the evolutionary root of requirements, recommended hods, and qualification procedures described in this appendix.
cific treatment of seismic qualification, part of the qualification test sequence recommended in E 323-1974, is addressed in IEEE 344-1987 (Reference 2). This appendix bases technical ance, recommendations, and requirements for seismic qualification on IEEE 344-1987.
AP1000 Equipment Qualification methodology addresses the expanded scope of E 627-1980 (Reference 3), which encompasses the qualification of Class 1E electrical and ty-related mechanical equipment. IEEE 627 generalizes the principles and technical guidance of E 323 and 344. Compliance with the IEEE 323-1974 and 344-1987 is the specific means of pliance with the intent of IEEE 627-1980 for safety-related electrical and mechanical equipment.
ety-related electrical and mechanical equipment is typically qualified using analysis, testing, or a bination of these methods. The specific method or methods used depend on the safety-related tion of the equipment type to be qualified. Safety-related mechanical equipment, such as tanks valves, is typically qualified by analysis, with supplementary functional testing when functional rability is demonstrated only through testing, as is the case for active valves. Either testing or ing combined with analysis is the method used for environmental and seismic qualification of ty-related (Class 1E) electrical equipment.
technical discussions of this appendix follow the format headings of the equipment qualification packages (EQDPs) to be issued as specific qualification program documentation. This atting (see Section 3D.7) permits easy cross-reference between the methodology defined in this ort and the detailed plans contained in the equipment qualification data packages. Attachment A is appendix is the format used for the equipment qualification data package.
chment B of this appendix, "Aging Evaluation Program," describes methods for addressing ntial age-related, common-mode failure mechanisms used in AP1000 equipment qualification grams. The approach conforms with current industry positions and makes maximum use of ilable data and experience in the evaluation, test, and analysis of aging mechanisms.
chment C, "Effects of Gamma Radiation Doses Below 104 rads on the Mechanical Properties of erials," provides the basis that radiation aging below 104 rads is not a significant factor in the ty of the equipment to perform properly during a seismic event. For some devices, electrical perties are degraded above 103 rads. Radiation aging for safety-related equipment which is ject to lifetime doses of less than 104 rads (103 rads for certain electrical components) and not ject to a high-energy line break environment is not required to be addressed in AP1000 lification programs.
chment D, "Accelerated Thermal Aging Parameters," describes the methodology employed in ulating the accelerated thermal aging parameters used in this program.
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1 Purpose basic objectives of qualification of safety-related electrical and mechanical equipment follow:
To reduce the potential for common cause failures due to specified environmental and seismic events To demonstrate that safety-related electrical and mechanical equipment is capable of performing its designated safety-related functions.
appendix describes the methodology that has been adopted to qualify equipment according to E 627-1980, "IEEE Standard for Design Qualification of Safety System Equipment Used in lear Power Generating Stations." The two standards primarily used to demonstrate compliance this standard are IEEE 323-1974, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear er Generating Stations," and IEEE 344-1987, "IEEE Recommended Practice for Seismic lification of Class 1E Equipment for Nuclear Power Generating Stations."
2 Scope qualification criteria, methods, and environmental conditions described constitute the hodology that is adopted to comply with the standards for the AP1000. This methodology applies afety-related, seismic Category I electrical and mechanical equipment and is also used for certain itoring equipment. Seismic Category II equipment is also within the scope of this program. The ria used for the design of seismic Category II structures, systems, and components are ussed in Section 3.7.
ormance during abnormal environmental conditions, while not specifically designated as an stry or a regulatory qualification requirement, is also addressed by this appendix. Performance ng normal service conditions is demonstrated by tests and inspections addressed by the ipment specification. Electromagnetic interference (EMI) testing or analysis is not included in the lification process and is addressed on an individual equipment basis, as necessary.
3 Introduction appendix identifies qualification methods used for the AP1000 to demonstrate the performance afety-related electrical and mechanical equipment when subjected to abnormal and accident ironmental conditions including loss of ventilation systems, feedline, steam line and main coolant em breaks, and seismic events. This appendix provides the expected conditions for various tions in the AP1000. General requirements for the development of plans/procedures/reports are provided. Section 3D.4 identifies the various industry and regulatory criteria upon which the gram is based. Section 3D.5 defines the design specifications and applicable test environments.
tion 3D.6 defines the basis for the qualification method selection. Section 3D.7 outlines the umentation requirements.
4 Qualification Criteria environmental requirements considered in the design of safety-related equipment are embodied DC 2, "Design Bases for Protection Against Natural Phenomena"; GDC 4, "Environmental and sile Design Bases"; and GDC 23, "Protection System Failure Modes." GDC 1, "Quality Standards Records," and Criterion III, "Design Control," Criterion XI, "Test Control," and Criterion XVII, 3D-2 Revision 1
qualification methods described in this appendix are used to verify the environmental design is and capability of the safety-related electrical and mechanical equipment supplied for the 000. The results of the verification, as well as the design basis for each equipment, is umented in an equipment qualification data package. (See Attachment A for sample format.)
ign control, test control, and quality assurance record keeping is performed through the AP1000 lity Assurance Program. (See Chapter 17.)
4.1 Qualification Guides E 323-1974 and 344-1987 serve as the basis upon which the AP1000 equipment qualification hodology demonstrates compliance with IEEE 627-1980. NRC regulations stated in CFR 50.49, "Environmental Qualification of Electrical Equipment Important to Safety for Nuclear er Plants," and NRC guidance provided in Regulatory Guide 1.89, and Regulatory Guide 1.100, orse IEEE 323-1974 and IEEE 344-1987, respectively. The intent of the more general E 627-1980 is addressed through conformance with IEEE 323 and 344.
4.1.1 IEEE Standards following lists additional standards and guides used in developing the methodology:
IEEE 98-1984, "IEEE Standard for the Preparation of Test Procedures for the Thermal Evaluation of Solid Electrical Insulating Materials" IEEE 100-1996, "IEEE Standard Dictionary of Electrical and Electronic Terms" IEEE 308-1991, "IEEE Standard Criteria for Class 1E Power System for Nuclear Power Generating Stations" IEEE 317-1983, "IEEE Standard for Electric Penetration Assemblies in Containment Structure for Nuclear Power Generating Stations" IEEE 381-1977, "IEEE Standard Criteria for Type Tests of Class 1E Modules Used in Nuclear Power Generating Stations" IEEE 382-1996, "IEEE Standard for Qualification of Actuators for Power-Operated Valve Assemblies with Safety-Related Functions for Nuclear Power Generating Stations" IEEE 383-1974, "IEEE Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations" IEEE 420-1982, "IEEE Standard Design and Qualification of Class 1E Control Boards, Panels, and Racks Used in Nuclear Powered Generating Stations" IEEE 494-1974, "IEEE Standard Method for Identification of Documents Related to Class 1E Equipment and Systems for Nuclear Power Generating Stations" IEEE 535-1986, "IEEE Standard for Qualifying Class 1E Lead Storage Batteries for Nuclear Power Generating Stations" 3D-3 Revision 1
IEEE 603-1991, "IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations" IEEE 649-1991, "IEEE Standard for Qualifying Class 1E Motor Control Centers for Nuclear Power Generating Stations" IEEE 650-1990, "IEEE Standard for Qualification of Class 1E Static Battery Chargers and Inverters for Nuclear Power Generating Stations" IEEE-741-1997, "IEEE Standard Criteria for the Protection of Class 1E Power Systems and Equipment in Nuclear Power Generating Stations" ANSI/IEEE C37.98-1987, "IEEE Standard for Seismic Testing of Relays."
4.1.2 NRC Regulatory Guides e area of seismic and environmental qualification of safety-related electrical and mechanical ipment, the NRC has issued the following Regulatory Guides:
ulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)" - The guide orses ANS and ANSI standards for quality assurance programs, but is considered here cifically for guidance in determining documentation adequacy. Appendix A of the guide, Item 9, cedures for Performing Maintenance," addresses procedural and documentation requirements maintenance of safety-related equipment, preventive maintenance, repair, and replacement. This e is a source in the development of qualification in the on-going qualification programs discussed ubsection 3D.6.4.
ulatory Guide 1.61, "Damping Values for Seismic Design of Nuclear Plants" - The guide cribes acceptable values of damping used in elastic modal dynamic seismic analysis of seismic egory I structures, systems, and components. The AP1000 equipment qualification program is ed on Regulatory Guide 1.61 and on values considered to be acceptable based on past NRC eptances. The safe shutdown earthquake (SSE) damping values used for the qualification of hanical and electrical equipment are listed in Table 3.7.1-1 of Chapter 3.
ulatory Guide 1.63, "Electric Penetration Assemblies in Containment Structures for Nuclear er Plants" - The guide endorses, with certain qualifications, IEEE 317-1983. External circuit ection of electric penetration assemblies should meet the provisions of Section 5.4 of E 741-1986, "Criteria for Protection of Class 1E Power Systems and Equipment in Nuclear erating Stations, as these are beyond the of scope IEEE 317. The AP1000 design complies with E 741-1997. The AP1000 equipment qualification program employs the recommendations of ulatory Guide 1.63, Revision 3, in specifying qualification plans as a means of supplementing the ance of IEEE 317 and 323.
ulatory Guide 1.73, "Qualification Tests of Electric Valve Operators Installed Inside the tainment of Nuclear Power Plants" - The guide endorses, with certain qualifications, E 382-1972. The AP1000 equipment qualification program employs recommendations of ulatory Guide 1.73, but gives preference to the guidance of IEEE 382-1996, where it is essary to supplement the guidance of IEEE 323 or 344 in specifying qualification plans for electric e operators.
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orsement of IEEE 323 is the reference to seismic qualification methods of IEEE 344 as a part of qualification test sequence. (See Regulatory Guide 1.100 later in this discussion.) The AP1000 ipment qualification methodology addresses the recommendations of Regulatory Guide 1.89 by following:
The recommendations of IEEE 323-1974 are met by the methods discussed in this appendix The radiation source terms used in qualification differ from those of Regulatory Guide 1.89, and are described in Section 3D.5 of this appendix The seismic qualification requirements employ the recommendations of IEEE 344-1987 as described in Attachment E of this appendix.
ulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic ponse Analysis" - The guide describes methods and procedures for the following:
Combining the values of the response of individual modes in a response spectrum modal dynamic analysis to find the representative maximum value of a particular response of interest for each of the three orthogonal seismic spatial components Combining the maximum values (or representative maximum values) of the responses for a given element of a system or item of equipment, determined for each of the three orthogonal spatial components.
AP1000 equipment qualification program employs methods consistent with the mmendations of Regulatory Guide 1.92 when combining individual modal response values or the onse of three independent spatial components in seismic analyses.
ulatory Guide 1.97, Revision 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants to ess Plant and Environs Conditions During and Following an Accident. The guide describes a hod acceptable to provide instrumentation to monitor plant variables and systems during and wing an accident in a light-water-cooled nuclear power plant. The AP1000 program, identified as post-accident monitoring instrumentation system (PAMS), provides the capability to monitor plant ables and systems operating status during and following an accident. PAMS includes those ruments provided to indicate system operating status and furnish information regarding the ase of radioactive materials.
ulatory Guide 1.100, "Seismic Qualification of Electrical Equipment for Nuclear Power Plants" -
guide endorses IEEE 344-1987. Regulatory Guide 1.100 particularly notes that IEEE 344-1987 pplied in the qualification of safety-related mechanical equipment, as well as Class 1E electrical ipment. The AP1000 equipment qualification methodology employs the recommendations of ulatory Guide 1.100, as described in Attachment E of this appendix.
ulatory Guide 1.122, "Development of Floor Design Response Spectra for Seismic Design of r-Supported Equipment or Components" - The guide describes specific methods for developing r (and other equipment mounting locations) response spectra. Included are specific criteria for the adening frequency amplitude peaks and smoothing of the frequency amplitude spectrum to rporate conservatism in the seismic requirements. This is to compensate for other uncertainties nalysis. The AP1000 equipment qualification program employs methods consistent with the mmendations of Regulatory Guide 1.122.
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cifying the qualification program plans where this guide supplements the guidance of IEEE 383 to further demonstrate conformance with the guidance of IEEE 323. As neither IEEE 383 nor ulatory Guide 1.131 specifically addresses considerations for cable field splices and nections, guidance for their qualification is taken from IEEE 572 and Regulatory Guide 1.156.
ulatory Guide 1.156, "Environmental Qualification of Connection Assemblies for Nuclear Power nts" - The guide endorses IEEE 572-1985. The AP1000 equipment qualification program loys the recommendations of Regulatory Guide 1.156 in specifying the qualification program s where this guide supplements the guidance of IEEE 572 to demonstrate conformance with the ance of IEEE 323.
ulatory Guide 1.158, "Qualification of Safety-Related Lead Storage Batteries for Nuclear Power nts" - The guide endorses IEEE 535-1986. The AP1000 equipment qualification program loys the recommendations of Regulatory Guide 1.158 in specifying the qualification program s where this guide supplements the guidance of IEEE 535 to demonstrate conformance with the ance of IEEE 323.
ulatory Guide 1.180, Guidelines for Evaluating Electromagnetic and Radio-Frequency rference in Safety-Related Instrumentation and Control Systems. Regulatory Guide 1.180 ides guidance to evaluate electromagnetic and radio-frequency interference in safety-related rumentation and control systems. The AP1000 equipment qualification program employs hods consistent with the recommendations of Regulatory Guide 1.180, where applicable.
ulatory Guide 1.183, Alternate Radiological Source Terms for Evaluating Design Basis idents at Nuclear Power Reactor. The radiation dose rates and integrated doses applicable for 000 following a design basis accident are determined based on the criteria of NUREG-1465 and regulatory guide.
4.2 Definitions nitions of terms used in this appendix are contained in the referenced standards and IEEE 100, e Authoritative Dictionary of IEEE Standard Terms, Seventh Edition." Subsection 3D.4.5 clarifies definitions of "life" (that is, design, shelf, and qualified life) as used in this methodology. The terms sign life" and "qualified life" have the meanings set forth in IEEE 323 and are used in the context at standard.
4.3 Mild Versus Harsh Environments lification requirements differ for equipment located in mild and harsh environments.
E 323 defines a mild environment as an environment expected as a result of normal service ditions and the extremes of abnormal service conditions where a safe shutdown earthquake is only design basis event of consequence or conditions where thresholds of material degradation reached. The following limits are established as the delimiting environmental parameter values mild and harsh environments.
ically a mild environment conforms with the environmental parameter limits of Table 3D.4-1, gh others may apply to specific equipment applications or locations.
scope of 10 CFR 50.49 is limited exclusively to equipment located in a harsh environment. The 000 equipment qualification program conforms with the requirements of 10 CFR 50.49 for the 3D-6 Revision 1
ain bounded by normal or abnormal conditions. Any equipment that is above 104 rads gamma for electronics) will be evaluated to determine if a sequential test which includes aging, radiation, the applicable seismic event is required or if sufficient documentation exists to preclude such a 4.4 Test Sequence ere the test sequence deviates from that recommended by IEEE 323-1974, the deviation is fied. The test sequence employed for a given hardware item is specified in the equipment lification data package Sections 2.1 and 3.6 (see Attachment A for example). Note that for this rence and subsequent references to Attachment A the information in Attachment A will be pleted in accordance with Subsection 3.11.5. Clarifications to the IEEE 323-1974 recommended sequence are discussed in the following:
Burn-In Test For electronic equipment, a burn-in test is completed, before operational testing of the equipment, to eliminate infant failures. The test consists of energizing the equipment for a minimum of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> at nominal voltage and frequency under ambient temperature conditions.
Any malfunction observed during these tests are repaired, and the 50-hour burn-in test is repeated for the repaired portion of the equipment.
Performance Extremes Test For equipment where seismic testing has previously been completed employing the recommended methods of IEEE 344-1987, seismic testing is not repeated. Testing of the equipment to demonstrate qualification at performance extremes is separately performed as permitted by IEEE 323-1974, Subsection 6.3.2(3). Additional discussion is provided in Subsection 3D.6.5.1.
Aging Simulation and Testing For equipment located in a mild environment, aging is addressed as described in Subsections 3D.6.3, 3D.6.4, and Attachment B. If there are no known aging mechanisms that significantly degrades the equipment during its service life, it is acceptable to perform seismic testing of unaged equipment. Separate testing or analysis (or both) is provided to demonstrate that the aging of components is not significant during the projected service or qualified life of the equipment.
Synergistic Effects An important consideration in the aging of equipment for harsh environment service is the possible existence of synergistic effects when multiple stress environments are applied simultaneously. This potential is addressed by conservatism inherent in the determination and use of the worst-case aging sequence and conservative accelerated aging parameters.
The combination of effects from pressure, temperatures, humidity, and chemistry are addressed by the high-energy line break (HELB) tests. Since the test item is not exposed to radiation during this test, the effects of this parameter are conservatively addressed by subjecting the test items to the required total integrated dose before the high-energy line break. Specifically for instruments, 3D-7 Revision 1
Visual Inspections/Disassembly The results of post-test visual inspections are not necessarily documented unless problems are discovered. Disassembly is performed only when test results or visual inspections require further investigation.
4.5 Aging 4.5.1 Design Life AP1000 equipment qualification program relies on the IEEE 323 definition for design life, icularly its distinction with respect to qualified life.
ead of determining a qualified life for mild environment equipment for which the seismic event is exclusive design basis event to be addressed, a design life is determined. Design lives offered in ufacturers' literature are accepted cautiously, particularly where the equipment is typically used applications outside the nuclear industry.
application of the design life is substantiated by sound bases in reliability theory and relevant stry standards, or experience data sources within the nuclear industry. Analyses treat the licability and similarity of the equipment and conditions relevant for the AP1000 safety-related lication. These analyses, and documentation of such, conform with guidelines of IEEE standards, pplicable, and with Sections 3D.6 and 3D.7 of this appendix.
4.5.2 Shelf Life ed on recommended storage environments, the shelf life of an equipment item is not typically a ificant portion of the defined qualified life. For example, ambient temperatures during storage are cally less than the operating temperatures assumed for aging calculations. Therefore, as long as ipment is in storage and is not energized (not experiencing self-heating), a reduction in qualified s not appropriate. However, if storage conditions differ significantly from those recommended or storage time becomes dramatically extended, the impact to the qualified life is determined by lication of the Arrhenius time-temperature relationship.
4.5.3 Qualified Life ualified life is established for each item of safety-related equipment that is exposed to a harsh ironment based on the conditions postulated at the equipment location with consideration of the ipment operability requirements.
determination of qualified life considers potential aging mechanisms resulting from significant ervice thermal, radiation, and vibration sources, and the effects of operational cycling chanical or electrical or both). Generally, all aging mechanisms do not apply to each item of ipment. Relevant aging mechanisms addressed or simulated are determined jointly with the tification of the equipment's critical components, functional modes, and material characteristics, the assessment of tolerable limits in degradation of the components. An a priori consideration in cting equipment to qualify is the evaluation of the equipment's inherent capability to survive and rate under the conditions for which it is qualified.
e past qualification tests have provided a substantial basis for this assessment (indeed, some provide sufficient basis to preclude any new testing as part of the AP1000 program) specific 3D-8 Revision 1
addressed in the documentation of qualification for each equipment type, as applicable.
lified life is established by the most limiting of the five aging mechanisms. Qualified life may be ed by the tolerable degradation of a single component or material critical to the equipment's ability to perform its safety function. Aging is subject to the requirement for margin. See section 3D.4.8 of this appendix.
some equipment, qualified life is established on the basis of periodic replacement of certain rt-lived, age-sensitive components. The user complies with the mandatory replacement practices umented in the equipment qualification data packages (see Subsection 3D.7.2.5 and chment A, Subsections 3.9.3 and 6.1) to affirm the equipment qualified life.
objective of thermal and irradiation qualified life testing is to simulate, according to the available irical material data, the degradation effects such that the equipment is in its end-of-life condition re the application of the design basis event conditions testing.
rmal qualified life is evaluated using the Arrhenius time-temperature relationship. (See more iled discussions in Attachments B and D of this appendix.) The activation energy is the exclusive erial-dependent parameter input into the Arrhenius time-temperature relationship. The activation rgy is an empirically determined parameter indicative of the thermal degradation of a physical perty of a material (for example, elasticity of silicone rubbers or insulation resistance of s-linked polyethylene cable insulation). Each material may have more than one physical property may be subject to thermal degradation over time. Consequently, it may have different activation rgies with respect to each property. Thus, the selection of activation energy considers the erial property most germane to the safety-related function of the material or component. (Also Subsection 3D.4.5.4.)
mon practice for the evaluation of irradiation-induced degradation is to consider the sum of mated life and the accident radiation doses before design basis event testing. When testing, the l dose is applied during the radiation aging simulation portion of the qualification test sequences.
is considered conservative because the equipment has accumulated an exposure, or total grated dose, before the initiation of the seismic and accident environment testing. Further bases est dose determination are provided in Subsection 3D.5.1.2. Sufficient margin must be included st parameters (see Subsection 3D.4.8). The same margins are applied in an analysis of radiation or design basis event radiation dosage.
simulation of age also includes the effects of operational cycling, both electrical and mechanical.
erally, these considerations are applied specifically to electromechanical equipment such as e operators, limit switches, motors, relays, switches, and circuit breakers. Furthermore, the ulation of these effects is waived where existing data demonstrates equipment durability greatly in ess of estimated number of operating cycles for Class 1E service. Analysis or justification is ided for any case where operational cycling is omitted in the test sequence.
not practicable to simultaneously simulate the aspects of aging. Development of each test plan siders known synergies and sequences the simulation of the various applicable aging hanisms with regard for conservatism of the overall effect on the test specimens.
4.5.4 Qualified Life Reevaluation ay be possible to extend the qualified life of a particular piece of equipment at some future date omparing the actual in-plant environments and conditions during the equipment's installed life to 3D-9 Revision 1
ding materials aging may be used. These efforts reveal the conservatism of the original thermal calculation, which assumes that the maximum value specified for the normal plant operating ironment endured at all times.
ough a strict Arrhenius calculation may yield an extended qualified life, care is taken in using this apolation because of uncertainties in the methodology. The Arrhenius time-temperature tionship relies on empirically determined activation energies of materials. This parameter has n determined for a number of materials to at least a good approximation for small temperature apolations. Extrapolation of the Arrhenius model to time periods of temperature beyond the range aterials test data is questionable and may result in large errors.
culated qualified lives based on this methodology should be limited to 20 years unless sound nical bases can be cited. This position is consistent with industry guidelines such as E 98-1984, NUREG/CR-3156 (Reference 4), and EPRI NP-1558 (Reference 5).
4.6 Operability Time post-accident operability times specified in Subsection 1.7.1 of each equipment qualification package (see Attachment A) are conservatively established based on the safety-related function ormed by that equipment for the spectrum of design basis event conditions. These include the wing:
Trip and/or monitoring functions of sensors and instruments Operability requirements for electromechanical equipment Duration of required operability for active valves.
evaluation also considers what consequences the failure of the device has on the operator's on or decisions and the mitigation of the event. Table 3D.4-2 lists and explains typical operability s.
monitoring functions, simulated aging techniques are employed to shorten the test time following gh-energy line break. These also comply with the margin guidelines of Subsection 3D.4.8.
gins for trip function requirements are contained in the high-energy line break envelopes that ompass a full spectrum of break sizes. The defined margins are also justified by the fact that the al generated by the sensor is locked in by the protection system and does not reset should the sor fail after completion of its designated trip time requirement.
4.7 Performance Criterion basic performance criterion is that the qualification test program demonstrates the capability of equipment to meet the safety-related performance requirements defined in the equipment lification data package, Section 1.7, while subjected to the environmental conditions specified in equipment qualification data package, Section 1.8. Where three or more specimens are tested, re of one of three may be considered a random failure, subject to an investigation concluding that observed failure is not indicative of a common-mode occurrence.
equipment for which aging is addressed by evaluation of appropriate mechanism(s) through a ew of available material and component information, the basic acceptance criterion is that the luation of test data demonstrates that the effect of aging is minor and does not affect the ability of the aged equipment to perform specified functions.
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ice conditions in order to establish the conditions for qualification. This margin is provided in er to account for normal variations in commercial production of equipment and for reasonable rs in defining satisfactory performance. Further guidance for determining the acceptability of gin with respect to application-specific or location-specific requirements is provided by the NRC UREG-0588 and Regulatory Guide 1.89, Revision 1. Margins are included in addition to servatisms applied during the derivation of the local environmental conditions of the equipment, ss the conservatism is quantified and specifically shown to meet or exceed the guidance of E 323, NUREG-0588, and Regulatory Guide 1.97.
sistent with IEEE 323, margin is incorporated into the specification of the generic qualification ameters by either increasing the test levels, number of test cycles, test duration, or a combination ese options as appropriate. The AP1000 generic qualification parameters are selected to elop a range of loss of coolant accident and high-energy line break sizes, and equipment tions. Margin in seismic conditions for test and analysis are addressed in Subsection 3D.4.8.4.
margins available for a specific application may be larger than the generic equipment lification test objective for seismic events and some events outside containment and are verified n application-specific basis.
efining qualification parameters, the AP1000 equipment qualification program incorporates gin as described in the following subsections. Table 3D.4-3 lists margin requirements applied.
generic testing, margin is applied at the time of testing to cover known safety-related applications e equipment. Generally, this results in a worst-case test that provides substantial margin for lications where lesser environments apply. Application of margin for seismic qualification resses several cases unique to the qualification approach. (See Subsection 3D.4.8.4.)
4.8.1 Normal and Abnormal Extremes ndicated in Section 7 of IEEE 323, the application of margin is directed at specifying adequate lification requirements for the most severe service conditions represented by the design basis nts (that is, high-energy line break accidents and seismic events). Consequently, the AP1000 ipment qualification methodology does not apply any systematic margin to the normal and ormal environment parameters in defining the qualification conditions.
electronic equipment not required to operate in a high-energy line break environment, additional gin is included by requiring that the equipment operate through the conservative normal and ormal service conditions indicated in Figure 3D.5-1. The environmental parameters at least equal specified range of service condition parameters. An exception occurs for transmitters where a ormance verification is completed at 130°F on each transmitter to encompass the specified imum abnormal conditions. For equipment to be qualified to operate in a high-energy line break ironment, qualification to the severe high-energy line break conditions demonstrates ample gin for acceptable performance under certain specified normal and abnormal service conditions.
4.8.2 Aging specific margin is applied to the time component in deriving appropriate aging parameters, if gin is included in deriving the accelerated aging parameters employed for simulating each licable aging mechanism.
gin may be addressed by demonstrating the adequacy of the aging simulated by test through the ulation of time-temperature equivalence (See Attachment B of this appendix) or the comparison 3D-11 Revision 1
ulations are subject to criteria, including the following:
Test temperature must endure for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Test temperature must exceed any application temperature (that is, the normal or abnormal environment in which the equipment is to be used, and for which the life is calculated)
Test temperature must be less than state-change temperature for materials critical to the equipment safety-related function or capability to endure the subsequent design basis event testing A conservative activation energy is used. Activation energies for materials critical to the equipment safety-related function or capability to endure the subsequent design basis event testing are considered. Materials may have several activation energies, each for a different material property. Relevant material properties are considered.
argin is not demonstrated through conservatism in the aging parameters or calculation, then a percent time margin is included.
argin of 10 percent in the other parameters (for example, irradiation, operational cycling) applies oth the aging simulation and the post-accident simulated aging, with few exceptions.
equipment required by design to perform its safety-related function within a short time period into design basis event (that is, within seconds or minutes), and having completed its function, sequent failure is shown not to be detrimental to plant safety, margin by percentage of additional or equivalent time-temperature is not applied. Margins for trip function requirements are tained in the worst-case high-energy line break envelope. Test parameters are simulated on a
-time basis with the transient condition margins listed in Table 3D.4-3. Trip signals, once erated by the sensors, are locked in by the protection system and do not reset in the event of sequent sensor failure.
4.8.3 Radiation additional 10 percent is added to the calculated total integrated dose in specifying the test uirements.
4.8.4 Seismic Conditions uired response spectra included in Subsection 3.7.2 or other AP1000 program specifications are conditions to be enveloped. No amplitude margin is added to these conditions. Peak broadening so discussed in Subsection 3.7.2. Seismic qualification by analysis addresses margin uirements by other methods of conservatism while using the same sets of requirements - no litude margin is included. For qualification tests, the test facility increases the amplitude of mic profiles by 10 percent to incorporate margin.
most applications, considerable margin exists with respect to the acceleration levels employed the width of the response spectra. Further details are addressed in Attachment E.
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lting from a spectra of loss of coolant accidents and high-energy line break sizes and locations, various nodes in the containment. As a consequence, these design envelopes already contain ificant margin with respect to any transient corresponding to a single break.
AP1000 equipment qualification methodology requires that the qualification envelopes be ved with a margin of 15°F and 10 psi with respect to the design envelopes in Figures 3D.5-2 and 5-3. The margin on dose is identified by comparing the location specific dose requirements and AP1000 equipment qualification parameters.
alkalinity of the chemistry is increased by 10 percent with respect to the peak value determined he AP1000 containment sump conditions.
4.9 Treatment of Failures primary purpose of equipment qualification is to reduce the potential for common mode failures to anticipated environmental and seismic conditions. The redundancy, diversity, and periodic ing of nuclear power plant safety-related equipment are designed to accommodate random res of individual components.
ere an adequate test sample is available, the failure of one component or device together with a cessful test of two identical components or devices indicates a random failure mechanism, ject to an investigation concluding that the observed failure is not common mode. Where fficient test samples prevent such a conclusion, any failures are investigated to ascertain whether failure mechanism is of common mode origin. Should a common mode failure mechanism be tified as causing the failure, either a design change is implemented to eliminate the problem or a eat test completed to demonstrate compliance with the criteria.
those mild environment equipment items that, through a review of available documentation, are ject to failure during a seismic event due to significant aging mechanisms, the material or ponent is replaced or monitored through a maintenance/surveillance program.
4.10 Traceability stem of baseline design documentation is instituted to control the design, procurement, and ufacturing of safety-related products. As part of this quality control program, critical parts are tified and assigned a level of control to reflect the estimate of potential qualification or urement problems. In addition, levels of quality inspection are also assigned to each part. The eline design documentation describes the equipment in sufficient detail (drawing number, part ber, manufacturer) to establish traceability between equipment shipped and equipment tested in qualification program.
4.10.1 Auditable Link Document purchaser of equipment referencing this program requires an auditable link document that ides a tie between the specific equipment and documentation of qualification reviewed for eptance under this program. This auditable link document includes one or more of the following ions, as applicable.
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reports. This link reflects a comparison of the as-built drawings, baseline design document or r documentation of the tested equipment to the specific equipment.
4.10.1.2 Component Link documentation certifies that the components (for example, replacement parts) used in the cific equipment are represented in the applicable test reports or via analysis under a component g program, such as that described in Attachment B (Subprogram B). This link applies only to ipment whose equipment qualification data package references a component testing program.
link reflects a comparison of the as-built drawings, baseline design document, or other umentation of the specific equipment to the component program listing.
4.10.1.3 Material Link documentation certifies that the materials used in the equipment are represented in a materials g analysis, such as that described in Attachment B, (Subprogram B). This link applies only to ipment whose equipment qualification data package references the materials aging analysis and cts a comparison of the as-built drawings, baseline design document, or other documentation of plant specific equipment to the materials aging analysis listing.
4.10.2 Similarity ere differences exist between items of equipment, analysis may be employed to demonstrate that test results obtained for one piece of equipment are applicable to a similar piece of equipment.
umentation of this analysis conforms with guidelines in IEEE 323 and 627, and section 3D.6.2.1 and Section 3D.7 of this appendix.
5 Design Specifications conditions and parameters considered in the environmental and seismic qualification of AP1000 ty-related equipment are separated into three categories: normal, abnormal, and design basis nt. Normal conditions are those sets and ranges of plant conditions that are expected to occur ularly and for which plant equipment is expected to perform its safety-related function, as uired, on a continuous, steady-state basis. Abnormal conditions refer to the extreme ranges of mal plant conditions for which the equipment is designed to operate for a period of time without special calibration or maintenance effort. Design basis event conditions refers to environmental ameters to which the equipment may be subjected without impairment of its defined operating racteristics for those conditions.
following subsections define the basis for the normal, abnormal, design basis event, and t-design basis event environmental conditions specified for the qualification of safety-related ipment in the AP1000 equipment qualification program. (These are cited in Section 1.7 of each ipment qualification data package; See Attachment A.)
service conditions simulated by the test plan are identified in equipment qualification data kage Section 3.7. (See Subsection 3D.7.4.6 and Attachment A.) In general, the parameters loyed are selected to be equal to (normal and abnormal) or have margin (design basis event and t-design basis event) with respect to the specified service conditions of equipment qualification package, Section 1.7, as recommended by IEEE 323. These conditions are conservatively ved to allow for possible alternative locations of equipment within the plant.
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ditions, as applicable, employing a cyclic test sequence of environmental and electrical extremes.
pical test profile, including voltage and frequency cycling, is shown in Figure 3D.5-1.
5.1.1 Pressure, Temperature, Humidity calculated values for temperature, pressure, and humidity during normal operation are specified able 3D.5-1 as a function of in-plant location.
5.1.2 Radiation Dose normal operating dose rates and consequent 60-year design expectation doses at various tions inside containment are specified in Table 3D.5-2. These values have been derived from retical calculations assuming an expected 60 years of continuous operation with a reactor power 468 MWth (including 2-percent power uncertainty) and steady-state operating conditions.
ivalent data at various locations outside containment are also specified in Table 3D.5-2.
total integrated dose employed for testing is a combination of normal and accident doses (where licable), and is defined to equal or exceed the maximum radiation dose contained in the ipment qualification data package. (See Section 3D.7 and Attachment A.) A margin of 10 percent cluded in defining the total integrated doses for testing. Normal operating and accident gamma es are simulated using a cobalt-60 or spent fuel source. The test dose is applied at a rate roximate to the maximum accident dose rate. Irradiation dose rates less than the maximum are sidered where there is significant shielding (greater than two mm of steel) or where the peak ontainment design basis event dose rate is not expected to affect the equipment's electrical ormance.
radiation dose rates encountered during normal operation for most equipment are not sidered critical parameters because of the resultant low total integrated dose (104 to 105 rads) ieved. For equipment not required post-accident, material can be selected based on previous test lts. Another test on the completed assembly is not required.
uipment is located in an environment where the normal total integrated dose exceeds the shold for radiation damage, then testing is required. For equipment required post-accident, the e received during normal operation is usually an insignificant part of the total integrated dose, uding accident conditions effects. The supposition that a concern over low dose rate effects inishes as the total integrated dose decreases is supported by Sandia National Laboratories tests ferences 6 and 7) on selected materials over a range of dose rates. These studies indicate that uction in original properties is about the same (and not significant) for dose rates up to a total grated dose in the megarad range. Although these tests were not performed at dose rates as low hose expected in a nuclear power plant and electrical properties were not evaluated, they do give e indication of the effect of varying the rate.
ed on results of research programs to date and low total integrated dose reached during normal ration, the AP1000 equipment qualification program does not consider degradation due to low e rate effects to be a significant concern. Therefore, the program does not include any action r than inspecting organic material degradation in the plant through normal maintenance.
5.2 Abnormal Operating Conditions ormal environments are defined to recognize possible plant service abnormalities that lead to rt-term changes in environments at various equipment locations.
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5.2.1 Abnormal Environments Inside Containment e AP1000 equipment qualification program there are multiple events postulated at least once r the 60 year design expectation which cause abnormal environmental conditions in the tainment. These are divided into two groups of events, based on peak containment temperatures ected.
up 1: 150°F Events Loss of a fan cooler Loss of all ac for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Pressurizer safety valve open/close during reactor coolant system transient.
up 2: 250°F Events Spurious automatic depressurization system (ADS) actuation Passive residual heat removal (PRHR) system use (long-term)
Reactor coolant system depressurization via pressurizer safety valve Small loss of coolant accident.
le 3D.5-3 presents the conditions associated with each of these abnormal environment events.
nt recovery occurs after each event with varying degrees of time and maintenance efforts. Thus, conditions resulting from these events are considered in the development of aging test ameters. Event frequency, conditions, and duration are accounted for within the context of the lified life objective of each equipment type test program.
5.2.2 Abnormal Environments Outside Containment re 3D.5-1 represents the assumptions made in defining potential abnormal environments due to of air-conditioning or ventilation systems.
le 3D.5-4 defines the abnormal environments as a function of equipment location. The assumed ation of the abnormal conditions specified in Table 3D.5-4 are consistent with operating practices technical specification limits. For certain plant applications, qualification for abnormal ironments is not necessary when equipment is located in environmental zones that do not exceed ufacturer's design limits for equipment operation.
5.3 Seismic Events Attachment E.
5.4 Containment Test Environment ulatory Guide 1.18 specifies that containment integrity is demonstrated at 1.15 times design sure. The design pressure of the AP1000 containment is 59 psig. Consequently, the maximum sure specified for the containment test is 59 x 1.15 = 67.85 psig. Other environmental ameters (such as temperature and humidity) of the containment test are adequately enveloped by parameters specified for normal or abnormal plant conditions.
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orms a safety-related function and which have a potential for changing the equipment ironment due to increased temperature, pressure, humidity, radiation, or seismic effects. The ironmental conditions for each applicable design basis event are summarized in Table 3D.5-5 and defined in the equipment qualification data package (see Section 1.8 of Attachment A) based on siderations and assumptions described in the following subsections.
5.5.1 High-Energy Line Break Accidents Inside Containment 5.5.1.1 Radiation Environment - Loss of Coolant Accident radiation dose rates and integrated doses following a design basis loss-of-coolant accident CA) are determined based on the criteria and guidance provided in NUREG 1465, Accident rce Terms for Light-Water Nuclear Power Plants - Final Report (Reference 8) and Regulatory de 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at lear Power Reactors (Reference 9).
radiation exposure inside the containment is conservatively estimated by considering the dose in middle of the AP1000 containment. Radioactive sources are assumed to be uniformly distributed ughout the containment atmosphere, and plate out of non-gaseous activity on containment aces is considered. No credit is taken for the shielding provided by internal structures and ipment.
rces are based on the emergency safeguards system core thermal power rating and the following lytical assumptions:
Power Level (including 2-percent power uncertainty) .................... 3,468 MWt Fraction of total core inventory released to the containment atmosphere:
Noble Gases (Xe, Kr)...................................................................................... 1.0 Halogens (I, Br)............................................................................................... 0.40 Alkali Metals (Cs, Rb) ..................................................................................... 0.30 Tellurium Group (Te, Sb, Se)........................................................................... 0.05 Barium, Strontium (Ba, Sr).............................................................................. 0.02 Noble Metals (Ru, Rh, Pd, Mo, Tc, Co) ........................................................... 0.0025 Lanthanides (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) ............................. 0.0002 Cerium Group (Ce, Pu, Np) ............................................................................ 0.0005 radionuclide groups and elemental release fractions listed above are consistent with the accident rce term information presented in NUREG-1465 and Regulatory Guide 1.183.
timing of the releases are based on NUREG-1465 assumptions. The release scenario assumed e calculations is described below.
nitial release of activity from the gaps of a number of failed fuel rods at 10 minutes into the dent is considered. The release of 3 percent of the core inventory of the volatile species (defined oble gases, halogens, and alkali metals) is assumed. An additional release period occurs over 3D-17 Revision 1
r the next 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, releases associated with an early in-vessel release period are assumed to ur, that is, from 40 minutes to 1.97 hours0.00112 days <br />0.0269 hours <br />1.603836e-4 weeks <br />3.69085e-5 months <br /> into the accident. This source term is a time-varying ase in which the release rate is assumed to be constant during the duration time. Additional ases during the early in-vessel release period include 95 percent of the noble gases, 35 percent e halogens, and 25 percent of the alkali metals, as well as the fractions of the tellurium group, um and strontium, noble metals, lanthanides, and cerium group as listed above.
re is no additional release of activity to the containment atmosphere after the in-vessel release se. Activity removal by natural mechanisms as described in Chapter 15 Subsection 15.6.5.3.2 Appendix B are considered only during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident.
above source terms are consistent with the guidance provided by the NRC in Regulatory de 1.183 for design basis accident (DBA) loss-of-coolant accident (LOCA) evaluations.
ed on these assumptions the instantaneous and integrated gamma and beta doses for the tainment atmosphere following a loss of coolant accident are shown in Figures 3D.5-2 and 5-3, respectively.
total integrated dose of radiation employed for testing is a combination of normal and design is event dose, as applicable. It is defined to equal or exceed the maximum radiation dose tained in the specification (Attachment A, Subsection 1.8.4.). A margin of 10 percent is included efining the total integrated dose for testing. Normal operating and design basis event gamma es are simulated using a cobalt-60 source. The test dose is applied at a rate approximate to the al phase of the design basis event dose rate shown in Figure 3D.5-2 as modified by shielding cts (typically 0.2 to 0.25 Mr/hr).
ere exposed organic material is evaluated by test for the effect of (accident) beta radiation, a beta rce is employed. Or a cobalt-60 or spent fuel source is used to impart the same dose using ma radiation. When doing beta equivalent testing, the total integrated dose using gamma is servatively equal to the beta total integrated dose, or the resulting bremsstrahlung is calculated the test item is exposed to an equivalent gamma dose.
iation conditions for loss of coolant accident envelop other scenarios, such as rod ejection.
5.5.1.2 Radiation Environment - Steam Line Break Accident rces associated with a steam line break accident are based on the release of reactor coolant em activity, assuming operation with the design basis fuel defect level of 0.25 percent. It is further umed that an event-initiated iodine activity spike occurs, which increases the reactor coolant vity during the accident based on a rate of increase that is 500 times the normal activity earance rate in the reactor coolant.
activity inventory is instantaneously released into the containment atmosphere. The dose is servatively estimated by considering the dose rate in the middle of the containment, with no credit he shielding provided by the internal structures, components, and equipment. The instantaneous integrated gamma and beta doses for the containment atmosphere following a steam line break shown in Figures 3D.5-4 and 3D.5-5, respectively.
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line break are equal to the values specified in Figures 3D.5-4 and 3D.5-5 for a steam line break.
5.5.1.4 Total Integrated Dose Specification applicable accident doses specified in equipment qualification data package Subsection 1.7.4 of chment A, have been derived based upon the time required to perform the specified safety tion in the accident environment (Attachment A, Subsection 1.6.1) and the dose calculations cribed previously, subject to the following modifications:
For equipment only required to provide trip or activation functions after accidents involving no release of radioactive material for at least one hour, the radiation dose is based on the normal dose rates (Table 3D.5-2).
5.5.1.5 Temperature/Pressure Environments design basis events addressed are the loss of coolant accident, steam line break and feedwater break. The WGOTHIC code is utilized to calculate the temperature and pressure conditions lting from these breaks. To retain the option of qualifying equipment for each of these high-rgy line break conditions, as applicable, separate environmental containment envelopes are cified for the higher irradiation/lower saturated temperature conditions of the loss of coolant dent as against the lower irradiation/short-term superheated temperature conditions associated the steam line break. To limit the number of basic envelopes, this latter envelope is servatively employed to define the containment environmental envelope following a feedline ak.
itionally, to facilitate AP1000 generic qualification and testing, the environmental envelopes have n combined to a single high-energy line break profile depicted in Figure 3D.5-8. This combined ile encompasses all locations inside containment on the basis of the containment analyses for the 000 design. The profile is used to qualify equipment for any application or location for the 000 consistent with the NRC requirements in 10 CFR 50.49 and IEEE 308, 323, 603, and 627 n margin is added and via conformance with IEEE 323 guidelines.
lification tests to high-energy line break conditions are designed to address the applicable cified environment(s) with a margin of 15°F and 10 psi. Separate envelopes with margin are loyed, or a combined loss of coolant accident/steam line break/feedwater line break envelope ures 3D.5-8 and 3D.5-9) may be employed for in-containment equipment qualification tests.
res 3D.5-8 and 3D.5-9 do not include margin from IEEE 323-1974, which will be incorporated in environmental qualification programs. The simulated post-design basis event aging
-temperature profile (Figures 3D.5-8 and 3D.5-9 from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to test conclusion) is defined sistent with the smallest value of activation energy applicable to the thermal aging sensitive ponents composing the test equipment or by a demonstrably conservative activation energy, as cribed in Attachment D.
5.5.1.6 Chemical Environment high-energy line break test will include chemical injection during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the test, to ulate the reactor coolant system fluid. Initial pH is from 4 to 4.5, with the solution consisting arily of boric acid.
e there is no caustic containment spray in the AP1000, subsequent adjustments in pH may not ecessary for all tests. Sump solution chemistry is adjusted by release of alkaline chemistry, 3D-19 Revision 1
gin in low pH value is not included, but is addressed by the continued injection through the first ours. Margin in alkaline pH, where adjustment is necessary, is incorporated by a 10 percent ease in alkalinity.
5.5.1.7 Submergence ormance of equipment in a submerged condition is verified by a test that replicates the actual ditions with appropriate margin.
5.5.2 High-Energy Line Break Accidents Outside Containment the majority of equipment located outside containment, the normal operating environment ains unchanged by a high-energy line break accident. As a consequence, qualification for such nts is covered by qualification for normal conditions.
mited amount of equipment located outside containment, near high-energy lines, could be subject cal hostile environmental conditions because of a high-energy line break outside containment. In case, the equipment is qualified for the conditions resulting from events affecting its location and which it is required to operate. Figure 3D.5-9 shows the design conditions for equipment that is uired to perform throughout postulated events. Figure 3D.5-9 does not include margin from E 323-1974, which will be incorporated in the environmental qualification programs. The imum pressure for any event outside containment is 6 psig.
6 Qualification Methods recognized methods available for qualifying safety-related electrical equipment are established EE 323. These are type testing, analysis, on-going qualification, or a combination of these hods. The choice of qualification method for a particular item of equipment is based upon many ors. These factors include practicability, size and complexity of equipment, economics, and ilability of previous qualification to earlier standards.
qualification method employed for each equipment type included under the AP1000 equipment lification program is identified in the individual equipment qualification data packages whether by (Attachment A, Section 3.0), analysis (Attachment A, Section 4.0), or by a combination of these hods. The AP1000 equipment qualification program may employ on-going qualification through use of maintenance and surveillance. Guidance for such an approach is not included in this endix.
6.1 Type Test preferred method of environmental and seismic qualification of safety-related electrical and tromechanical equipment for the AP1000 equipment qualification program is type testing ording to the guidelines and requirements of IEEE 323-1974 and 344-1987. Development of type requirements are discussed in Section 3D.5. Documentation requirements and test plan elopment are addressed in Section 3D.7.
itionally, qualification based on type tests performed according to IEEE 323 and 344, but not cifically for the AP1000, may be used as a qualification basis. Subsection 3D.6.5 of this appendix usses the combination of qualification methods as they apply to the AP1000 equipment lification program. (See Subsection 3D.6.5.1.)
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primary requirement is the demonstration of structural integrity during a seismic event. For ipment that performs an active or dynamic function, seismic qualification by analysis may also be
- d. (See Section E.3 of Attachment E.) However, the similarity between a qualified test unit and an upplied unit must be demonstrated unless otherwise justified. Subsection 3.9.2.2 describes the lification requirements for safety-related mechanical equipment where a fluid pressure boundary volved. For those mechanical components that are not pressure boundaries, analysis is ormed in compliance with the applicable industry design standard. Where age-sensitive erials, such as gaskets and packing, are used in the assembly of mechanical equipment, the g of these materials is normally evaluated based on an item-by-item review of the aging racteristics of the material. (See Subsection 3D.6.2.3.)
uirements for documentation of the analysis are further treated in Section 3D.7.
6.2.1 Similarity ilarities among manufacturer's models provides several options for extending qualification to ipment without the need for a complete qualification test program.
odel series, such as that for a solenoid valve design, consists of numerous models that are tical in materials of construction and manufacturing process, but have minor variance in size, tional mode, operating voltage, electrical termination type, and mechanical interface sizing. Such ances in most cases have no impact on or relevance to the capability of the various models to orm acceptably under environmental or seismic (or both) qualification test conditions.
hermore, the design basis document may apply equally to each member of the model series. In h cases, all members of the model series can be qualified by reference to the same testing or lysis.
re may be sufficient similarities between different model series to justify the case for similarity. A umented comparison addressing differences in the design for each, or apparent physical rences between members of each model series, may be sufficient to preclude the testing of one el series based on the testing of the other.
ilarly, different models of a manufacturer's transmitters may be identical in some respects but rent in others. The justification of similarity addresses the degree of similarity for critical racteristics. Differences that are not significant to qualification are also addressed for pleteness. The mechanical and electrical functional modes and configurations must be the
- e. The materials of construction may be different, but must demonstrate equivalent performance.
er means of assuring accuracy may be necessary. When the devices are sufficiently similar in all butes affecting qualification, qualification testing of one item can adequately cover another.
6.2.2 Substitution objectives are to establish a degree of similarity and equivalence of performance for parts and erials that are different and, ultimately, to preclude the need for testing. For example, a gasket erial is changed or a new type of capacitor is used because the original is no longer available, nomical, or inadequate. Substitution of parts and materials is acceptable if comparison or analysis ports the conclusion that equipment performance is the same or better as a result. Consideration ven to characteristics of materials and the relative degree to which each is affected (or degraded) he environmental parameters of qualification.
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e failures due to environmental effects of a design basis accident. Requirements are based on C 4 and 10 CFR 50, Appendix B. These criteria mandate that safety-related structures, systems, components be designed to accommodate both normal and accident environmental effects.
6.2.3.1 Equipment Identification ety-related mechanical equipment to be qualified is identified through the review of design basis umentation or the requirements of each safety-related fluid system. Only nonmetallic parts or components within the safety-related mechanical equipment are addressed for the effects of the tulated environments. The principal scope is typically valve "soft parts" that are critical to the valve ty-related function or pressure boundary integrity.
types of components most frequently encountered in the mechanical equipment evaluations are ussed in Subsection 3D.6.2.3.3. Properties of materials that are assessed to provide confidence afety-related function performance are also identified.
6.2.3.2 Safety-Related Function ety-related functions and performance criteria are identified based on system and component sification. Structure, system, and component design basis documentation is reviewed to rmine the specific safety functions. Components and subcomponents not involved in the ipment's safety-related function(s) are excluded from the qualification process if it is shown that r failures have no effect on the safety-related functions.
6.2.3.3 Performance Criteria prehensive performance criteria are established to satisfy the fundamental qualification uirements. The criterion for qualification is that the property of the nonmetallic material with regard s application is not degraded during the specified qualified life to the point that the component is ble to perform its intended safety-related function. Properties for the component types listed in le 3D.6-1 are discussed as examples.
kets and O-Rings capability of gaskets and O-rings to keep their shapes determines their ability to maintain sure boundaries. When an O-ring or gasket loses its dimensional memory, it does not exert the essary force on the confining surfaces. This could result in leakage. Compression set and gation are good indicators of the dimensional memory of a material. They also reflect the extent ermal aging and radiation-induced cross-linking. A compression set of 50 percent is chosen as a servative end-of-life criterion even though leakage is unlikely to occur until the component takes a pression set of greater than 75 percent. When compression set data is not available for a gasket
-ring, elongation at break is the material property evaluated because like compression set, it is ndication of dimensional memory and cross-link.
phragms phragms must remain flexible yet maintain their dimensional memory throughout the estimated hanical cycles. Retention of elongation or tensile strength is evaluated for radiation and thermal g.
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g to diaphragm support materials is retention of elongation.
ricants of the primary functions of oils and greases is to maintain a thin film barrier between moving s to reduce friction and wear. Irradiation reduces the capability of a lubricant to perform this tion by decreasing viscosity in oils and increasing penetration in greases and finally converting icants to hard, brittle solids if exposure is severe.
rm Gears m gears must be capable of transmitting forces without excessive deformation. Flexural strength e material property chosen to evaluate radiation and thermal aging resistance of worm gears.
6.2.3.4 Identification of Service Conditions vice conditions are identified for the normal and accident conditions. The general design of ipment permits exemption of environmental parameters such as pressure and humidity. Where cal parts are totally enclosed by metal and not directly exposed to potentially harsh environments, effects of humidity and chemical spray are not addressed. The degradation of mechanical ipment due to thermal and radiation aging is typically more severe than the possible degradation to other environments. Since most mechanical equipment interfaces with process fluid, the effect e fluid on the environmental conditions (temperature, radiation, and chemical) is considered.
6.2.3.5 Description of Potential Failure ere applicable, potential failure modes are identified and assessed for the equipment.
essment of equipment aging mechanisms is essential to determine if aging has a significant ct on operability. This assessment provides confidence that significant aging mechanisms are kely to contribute to common-mode failures adverse to the safety-related function of equipment.
6.2.3.6 Qualification Procedure nonmetallic materials identified are evaluated to the normal and accident environmental ameters. The evaluation procedure includes the following steps:
Identification of the environmental effect on the material properties Performance of a thermal aging analysis Determination of the environmental effects on the equipment safety-related function.
se are detailed in the equipment qualification data package of Attachment A, Section 4.Y.
6.2.3.7 Performance Criteria nonmetallic subcomponents of the mechanical equipment:
are acceptable for the plant environment by exhibiting threshold radiation values above the postulated environmental condition, and 3D-23 Revision 1
does exhibit a service life sufficient to survive the accident duration, or instead of a, b, and c, are acceptable for the plant environment by analysis that demonstrates that the safety-related function of the component is not compromised.
mechanical equipment is considered qualified if subcomponents important to the safety function acceptable.
metallic subcomponents not meeting the criteria must have a replacement interval specified to ntain the qualification of the affected equipment. The replacement interval is determined by lysis and documented.
6.2.3.8 Equipment Qualification Maintenance Requirements maintenance requirements resulting from the activities described herein are identified. The lification maintenance requirements are based on the following:
Qualification evaluation results (for example, periodic replacement of age-susceptible parts before the end of their qualified lives)
Equipment qualification-related maintenance activities derived from the qualification report(s)
Vendor recommended equipment qualification maintenance. Vendor recommended maintenance is included if it is required in order to maintain qualification.
6.2.3.9 Qualification Documentation qualification of the mechanical equipment to the postulated environments is documented in an itable form. See Section 3D.7.
6.3 Operating Experience lification by experience is not employed in the AP1000 equipment qualification program as a hod of qualification.
6.4 On-Going Qualification AP1000 equipment qualification program may employ on-going qualification through special ntenance and surveillance activities. However, this method of qualification is not suitable as a means for qualifying equipment for design basis event conditions. On-going qualification, as a hod, is used exclusively for safety-related equipment located in a mild environment area. Such requires supplementary test, or analysis to address equipment operability and performance ng and after a seismic design basis event.
umentation requirements for qualification that includes on-going qualification as a method are eloped to conform with NRC guidance provided in Regulatory Guide 1.33, Revision 2.
6.5 Combinations of Methods lification by a combination of the preceding methods may be used under the AP1000 equipment lification program.
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guidelines of IEEE 344-1987 and the environmental qualification program satisfies the guidelines EEE 323-1974.
lification test and analysis reports conforming to those IEEE Standards, but not specifically ormed to the AP1000 equipment qualification program parameters, may be acceptable as lification bases. In such cases, supplementary qualification efforts described in sections 3D.6.2, 3D.6.3, and 3D.6.4 of this appendix may be required to validate acceptability er the AP1000 equipment qualification program. Justifications are documented as analyses, and ear in equipment qualification data package, Section 4.0. (See Attachment A.)
6.5.1.1 Aging t qualification tests may provide sufficient basis to preclude new aging simulation testing as part e AP1000 program. Also, simulation of both electrical and mechanical operational cycling may waived where existing data demonstrates equipment durability greatly in the excess of the mated number of operating cycles for Class 1E service. Application of past qualification and other s is considered in the development of test plans and analysis procedures. The bases and fication is provided in qualification documentation for cases where applicable aging parameters omitted from the test sequence.
6.5.1.2 Seismic mic qualification generally relies on analyses and justification to verify the adequacy or licability of generic testing to a particular installed configuration of similar equipment. Analytical hods and documentation guidelines of IEEE 344-1987, as supplemented by Regulatory de 1.100, Revision 2, address these needs. Attachment E of this appendix provides the AP1000 ipment qualification program requirements regarding seismic qualification.
6.5.1.3 High-Energy Line Break Conditions ically, existing qualification tests address conditions of high-energy line break environments urring inside containment. These are used where it is demonstrated that the qualification elops the applicable requirements.
7 Documentation AP1000 equipment qualification program documentation consists principally of three types of uments:
"Methodology for Qualifying AP1000 Safety-Related Electrical and Mechanical Equipment" is the generic program "parent" document. It describes the methods and practices employed in the AP1000 equipment qualification program.
Equipment qualification data packages are "daughter" documents to the methodology. Each is a summary of the qualification program for a specific equipment type (for example, a particular model or design series of a manufacturer, an as-provided system, or a family of equipment tested as a set). The equipment qualification data package defines the qualification program objectives, methods, applicable equipment performance specifications, and the qualification plan. It provides a summary of the results.
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equipment qualification data packages are developed separate from the parent document.
ilarly, the equipment qualification test reports are developed separate from the equipment lification data packages. Equipment qualification test reports used in the AP1000 equipment lification program may include existing reports of testing or analysis that comply with the relevant ects of this methodology. Information necessary to demonstrate the equipment's capability to orm its intended safety-related function(s) while exposed to normal, abnormal, accident, and t-accident environments is provided in or referenced by the equipment qualification data package.
aintenance, refurbishment, or replacement of the equipment is necessary to provide confidence e equipment's capability to perform its safety function, this information is also included in the ipment qualification data package. Data, in raw form, cited in the equipment qualification data kages or equipment qualification test reports is available for audit for the life of the plant.
7.1 Equipment Qualification Data Package chment A contains sample of the equipment qualification data package format. Each equipment lification data package consists of the following elements:
tion 1.0 - Specifications tion 2.0 - Qualification Program tion 3.0 - Qualification by Test tion 4.0 - Qualification by Analysis tion 5.0 - Qualification by Experience (Not Used) tion 6.0 - Qualification Program Conclusions le 1 - Qualification Summary following paragraphs discuss the six sections in the equipment qualification data packages.
7.2 Specifications tion 1.0 of the equipment qualification data packages (Attachment A) contains the performance cification of the equipment. This specification establishes the necessary parameters for which lification is demonstrated. The basic criterion for qualification is that the safety-related functional uirements defined in Section 1.0 are successfully demonstrated, with margin, under the specified ironmental conditions.
following sections define the bases on which the parameters contained in Section 1.0 are cted.
7.2.1 Equipment Identification ipment is identified in Section 1.1 of Attachment A by manufacturer, model or model series, and rence to other documents describing or depicting its construction, configuration, and ifications that are uniquely necessary after manufacture to its application in the AP1000 plant ign. Model series (for example, a limit switch design family) and other pertinent details on items ing up the equipment type qualified are compiled as a table and referenced from this section.
7.2.2 Installation Requirements hat the qualification represents the in-plant condition, the method of installation, as specified in tion 1.2 of Attachment A, is in accordance with the supplier's installation instructions. Differences 3D-26 Revision 1
7.2.3 Electrical Requirements pertinent electrical requirements are specified (for example, voltage, frequency, load) in this ion. Also included is any variation in the defined parameters for which the equipment is to orm its specified functions (Section 1.3 of Attachment A).
7.2.4 Auxiliary Devices etimes the equipment qualified relies upon the operation of auxiliary devices in order to perform specified safety-related functions. These devices are identified in Section 1.4 of Attachment A.
iliary devices include items such as electrical conductor seal assemblies that, in service, become of the qualified equipment's pressure boundary. The applicable equipment qualification data kage for the auxiliary device(s) is specified, if known.
7.2.5 Preventive Maintenance ventive maintenance (Section 1.5 of Attachment A) to be performed includes maintenance or odic activities assumed as part of the qualification program or necessary to support qualification.
y those activities that are required in order to support qualification or the qualified life are cified. The manufacturer's recommended maintenance activities are considered to determine that e is no adverse impact to qualification or the maintenance of qualified life. Likewise, ufacturer's recommendations for maintenance or surveillance activities necessary to support rability are identified, or reference is made to the appropriate technical manual or supplements.
ne" means that maintenance is not essential to qualification or the qualified life of the equipment.
ever, this should not preclude development of a preventive maintenance program designed to ance equipment performance and to identify unanticipated equipment degradation as long as h a program does not compromise the qualification status of the equipment. Surveillance vities may also be considered to support a basis for and a possible extension of the qualified life.
7.2.6 Performance Requirements tion 1.7 of Attachment A contains a tabulation of performance requirements for each safety-ted function for which the equipment is qualified. Several such sections or tables may be essary when the equipment is qualified for applications where the performance requirements
. Performance requirements are stated regarding the normal and abnormal environmental ditions applicable at the location where the equipment is installed. Similarly, each design basis nt and the subsequent post-event period is included in the table.
gin is not included in the performance requirements except by conservatism in their rmination.
7.2.7 Environmental Conditions hin each set of performance requirements, a set of environmental parameters is specified in ion 1.8 of Attachment A, also in tabular form. Parameters are based on the equipment location function and include those addressed in other sections of this appendix.
gin is not included in the environmental parameters except by conservatism in their rmination. The objective is to provide the baseline reference onto which margin is added.
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tion 2.0. Attachment A includes a table to be completed as a graphic reference. As it is assumed tests, analyses, or some combination of the two are the principal methods of qualification, mns are included for each. Other methods, when used, are summarized in brief notes appended e table.
erences to reports of testing, analysis, or other information considered in support of the lification program are compiled in Section 2.2 of Attachment A. This includes any technical uals, drawings, and supporting material cited or referenced by text throughout the equipment lification data package.
7.4 Qualification by Test lification by test is selected as the primary method of qualification for complex equipment not dily amenable to analysis or for equipment required to perform a safety-related function in a high-rgy line break environment. The proposed test plan is identified in Section 3.0 of Attachment A.
ere supportive analysis is claimed as an integral part of the qualification program, cross reference ovided to Attachment A, Section 4.0 for those aspects of the qualification not covered by the test
. The following sections establish the basis on which the information specified in Section 3.0 is cted.
7.4.1 Specimen Description equipment qualified is identified, including the baseline design document number/reference, re applicable, the equipment type, manufacturer and model number, in Section 3.1 of chment A. When testing a model series (or equipment families), the representative items tested clearly identified. The basis of their representation should be included.
tion 3.1 is primarily intended to identify test specimens used in a test supporting the qualification gram. But it also discusses the specimens considered for other methods used in the qualification gram.
7.4.2 Number Tested test program is based upon selectively testing a representative number of components ording to type, size, or other appropriate classification, on a prototype basis. The number of items quipment representative of the equipment type that are tested is defined in Section 3.2 of chment A.
7.4.3 Mounting method of mounting the equipment for the test is identified in Section 3.3 of Attachment A. The lant installation requirements, as specified by the supplier under Section 1.2 of Attachment A, are represented.
7.4.4 Connections equipment connections necessary to demonstrate safety-related functional operability during ing are identified in Section 3.4 of Attachment A. This includes items that are part of the installed figuration, but are not part of the test apparatus.
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ecially under aggressive or adverse environmental conditions. Their thermal degradation and sitivity to irradiation and chemistry environments are considered in the qualification program, both mpact to equipment performance under harsh conditions and for their contribution to equipment lified life.
7.4.5 Test Sequence preferred test sequence specified in Attachment A, Section 3.5 is the one recommended by E 323-1974. The qualification test sequence used is specified in Section 3.6 of Attachment A.
ification for departures or additions to the preferred test sequence are included. Also, any portion e test sequence that is supplemented by analysis or other methods is identified for pleteness.
7.4.6 Simulated Service Conditions service conditions simulated by the test plan are identified in Attachment A, Section 3.7. In eral, the parameters employed are selected to be equal to (normal and abnormal) or have margin ident and post-accident) with respect to the specified service conditions of Attachment A, tion 1.8. Criteria for margin is detailed in Subsection 3D.4.8.
7.4.7 Measured Variables parameters measured during the specified test sequence in order to demonstrate qualification he performance specification (Attachment A, Section 1.0) are individually listed in Attachment A, tion 3.8 of Attachment A. This section is formatted to include parameters relevant to the test ironment and the electrical and mechanical characteristics of equipment operation. Other racteristics unique to a particular test or equipment type are included, when applicable.
7.4.8 Type Test Summary tion 3.9 of Attachment A provides a narrative summary of the qualification tests and results. The licable test reports are provided as references in Attachment A, Section 2.2. Test data is available audit throughout the operation of the plant.
h test report referenced by the equipment qualification data package should contain information d in the preceding section, as well as the following:
The test facility, location, and a description of the test equipment used. Monitoring equipment should have current calibration traceable to the National Bureau of Standards.
Test setup and specimen installation details.
Description of the mounting conditions simulated during the test program and any difference between them and the mounting details shown on the equipment drawings, with qualification of any differences found.
Description of limitations on the use and mounting of the qualified equipment found as a result of the qualification test program.
Description of the test method and the justification that the method meets the specification test requirements.
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Test records (for example, test response spectra, time history; accident transient parameters - temperature, pressure). This includes performance and operability test results, inspection results, and the monitored test and specimen and calibration records of instruments used.
Record of compliance of test results with the seismic qualification criteria.
Description of anomalies found during the test program, and their resolution(s).
ential aging mechanisms resulting from significant in-service thermal, electrical, mechanical, ation, and vibration sources are identified in Subsection 3.9.3 of Attachment A. When aging is ressed as part of the test sequence, the method employed for aging the equipment is indicated is chosen to conservatively simulate the potential aging effects resulting from the operating es and environmental conditions specified in Attachment A, Section 1.0. The methods employed ddress each of the potential aging mechanisms are discussed.
7.5 Qualification by Analysis lification by this methodology does not rely solely on analyses. Generally, analysis is permitted to port qualification testing or to establish that testing of other sufficiently similar equipment can be d to establish or extend the qualification of equipment covered by the equipment qualification data kage.
sample format for Section 4.0 of Attachment A is formatted to conform with the mmendations of IEEE 323-1974. Each subsection addresses a particular analysis if more than is performed to support qualification. Not all subsections identified in the sample format apply to particular analysis. Documentation of analyses demonstrating or supporting seismic qualification forms with the guidelines of Attachment E and the recommendations of IEEE 344-1987.
7.6 Qualification by Experience method of qualification is not used.
7.7 Qualification Program Conclusions tion 6.0 of Attachment A summarizes the conclusions of the qualification program, including and ressing methods employed and conditions upon which qualification of the equipment is based.
ails regarding each aspect of simulated aging are addressed distinctly, with conclusions as to the limiting aspects clearly stated.
clusions for each design basis event are summarized. Generally, these are combined as either ign basis event seismic and design basis event environmental.
7.8 Combined License Information used.
8 References IEEE-323-1974, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations."
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IEEE-627-1980, "IEEE Standard for Design Qualification of Safety System Equipment Used in Nuclear Power Generating Stations."
NUREG/CR-3156, "A Survey of the State-of-the-Art in Aging of Electronics with Application to Nuclear Plant Instrumentation."
EPRI NP-1558, Project 890-1, "A Review of Equipment Aging Theory and Technology."
NUREG/CR 2156, "Radiation Thermal Degradation of PE and PVC: Mechanism of Synergism and Dose Rate Effects," Clough and Gillen, June 1981.
NUREG/CR 2157, "Occurrence and Implication of Radiation Dose Rate Effects for Material Aging Studies," Clough and Gillen, June 1981.
NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants - Final Report," L. Soffer, et al., February 1995.
Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.
e: See Subsection 3D.4.1.1 for other IEEE references.
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Parameter Limit Notes perature 120°F ssure Atmospheric (Nominal) midity 30 - 65% (Typical) 95% (Abnormal) iation 104 rads gamma 103 rads gamma (IC electronics and microprocessors) mistry None mergence None 3D-32 Revision 1
Equipment Required Post-Accident Operability ipment necessary to perform trip 5 minutes (Envelops trip time requirements) tions ipment located outside containment, 2 weeks ccessible, and can be repaired, aced, or recalibrated ipment located inside containment 4 months (This number is based on an acceptable is inaccessible and is required for amount of time to be repaired, replaced, or t-accident monitoring recalibrated, or for an equivalent indication to be obtained.)
ipment located inside containment, is 1 year cessible, or cannot be repaired, aced, recalibrated or equivalent cation cannot be obtained ipment in a location that will have a Various (Specific as to function, maximum of 1 year) environment following an accident or ipment that does not provide rmation for a Type A, B, or C primary t-accident monitoring parameter 3D-33 Revision 1
Required Condition Parameter Margin Notes RMAL: Aging +10% +10% time margin, +10% radiation and/or selection of conservative test parameters. Comply with guidance of Subsection 3D.4.8.2 NORMAL: Temperature/ Margin is in "time" at abnormal test extremes.
Humidity Pressure None Nominally atmospheric.
Radiation +10% Include in aging doses, if applicable.
Voltage & +/- 10% Simulated during temperature/humidity test.
Frequency CIDENT: Transient Temperature (+15°F) and pressure (+10 psig peak)
Temperature and margins added to transient profile.
Pressure Chemical effects +10% In alkalinity of adjusted sump pH. Not applicable outside containment.
Radiation +10% Added to calculated total integrated dose.
Submergence Note 1 Generally, precluded by design.
Seismic/ +10% Of acceleration at equipment mounting point for either Vibration SSE or line-mounted equipment vibration. (See Subsection 3D.4.8.4.)
Post-accident +10% In time demonstrated via Arrhenius time/temperature Aging relationship calculation.
Margin in submergence conditions is achieved by increases in the post-accident time duration (+10%) and chemistry (+10% in alkalinity of adjusted sump pH).
3D-34 Revision 1
(Notes 1 and 2) ation/Parameter Normal Range Notes e 1 - Containment om numbers: 11000 through 11999)
Temperature 50° - 120°F Pressure -0.2 - +1.0 psig Humidity 0 - 100%
Radiation see Table 3D.5-2 Chemistry None e 2 - Auxiliary Building - Non-Radiological - I&C, DC Equipment, RCP Switchgear & Battery rooms, etc.
om numbers: 12101, 12102, 12103, 12104, 12105, 12111, 12112, 12113, 12201, 12202, 12203, 12204, 12205, 07, 12211, 12212, 12213, 12301, 12302, 12303, 12304, 12305, 12311, 12312, 12313, 12405, 12411, 12412, 01, and 12505)
Temperature 67 - 77°F (All rooms except 12405 and 12505) 50 - 85°F (Rooms 12405 and 12505)
Pressure Slightly positive to slightly negative Humidity 10 - 60%
Radiation <103 rads gamma Chemistry None e 3 - Auxiliary Building - Non-Radiological - Main Control Room om number: 12400, 12401)
Temperature 67 - 78°F Pressure Slightly positive Humidity 25 - 60%
Radiation <103 rads gamma Chemistry None e 4 - Auxiliary Building - Non-Radiological - Accessible om numbers: 12321, 12421, 12422, 12423)
Temperature 50 - 105°F Pressure Slightly positive Humidity 10 - 60%
Radiation <103 rads gamma Chemistry None 3D-35 Revision 1
(Notes 1 and 2) ation/Parameter Normal Range Notes e 5 - Auxiliary Building - Non-Radiological - MSIV Compartments om numbers: 12404, 12406, 12504, 12506)
Temperature 50 - 130°F Pressure Atmospheric Humidity 10 - 100%
Radiation <104 rads gamma Chemistry None e 6 - Auxiliary Building - Radiological - Inaccessible om numbers: 12154, 12158, 12162, 12163, 12166, 12167, 12171, 12172, 12254, 12255, 12256, 12258, 12262, 64, 12265, 12354, 12362, 12363, 12365, 12371, 12372, 12373, 12374, 12454, 12462, 12463)
Temperature 50 - 130°F Pressure Slightly negative to atmospheric Humidity 10 - 100%
Radiation See Table 3D.5-2 Chemistry None e 7 - Auxiliary Building - Radiological - Accessible om numbers: 12151, 12152, 12153, 12155, 12156, 12161, 12169, 12241, 12242, 12244, 12251, 12252, 12261, 68, 12271, 12272, 12273, 12274, 12275, 12341, 12351, 12352, 12361, 12451, 12452, 12461, 12553, 12554, 55, 12561)
Temperature 50 - 104°F Pressure Atmospheric Humidity 10 - 100%
Radiation See Table 3D.5-2 Chemistry None e 8 - Turbine Building om numbers: 20300 through 20799)
Temperature 50 - 105°F Pressure Atmospheric Humidity 10 - 100%
Radiation <103 rads gamma Chemistry None 3D-36 Revision 1
(Notes 1 and 2) ation/Parameter Normal Range Notes e 9 - Auxiliary Building - PCS Valve Room om number: 12541, 12701)
Temperature 50 - 120°F Pressure Atmospheric Humidity 10 - 100%
Radiation See Table 3D.5-2 Chemistry None e 10 - Auxiliary Building - Non-Radiological - Valve/Piping Penetration Room with SG Blowdown om number: 12306)
Temperature 50 - 105°F Pressure Slightly positive Humidity 10 - 60%
Radiation <103 rads gamma Chemistry None e 11 - Auxiliary Building - Radiological - Fuel Handling Area om numbers: 12562, 12563, 12564)
Temperature 50 - 105°F Pressure Slightly negative Humidity 10 - 100%
Radiation See Table 3D.5-2 Chemistry None s:
Room numbers - see Section 1.2, General Arrangement drawings.
Relative humidity is not controlled except in the main control room.
3D-37 Revision 1
Gamma Dose Rate 60-Year Gamma Dose Location (Rad air hour) (Rads air) de Containment:
CS Pipe - Center 1.9x103 1.0x109 CS Pipe - ID 1.1x103 5.7x108 CS Pipe - OD (contact) 7.8x101 4.1x107 CS Pipe - General Area(b) 4.0x101 2.1x107 utside Loop/Compartment Wall <0.1 <5x104 utside CA01 Excluding Rooms 11104 and <0.45 <2.4x105 1204 djacent to Reactor Vessel Wall 3.6x104 1.9x1010(a) side Containment:
enetration Area -- <2x107 ump Cubicles -- <2x107 adioactive Waste Area -- <2x107 adwaste Tank Cubicles -- <5x107 ther General Areas Not Under Radiation -- <1x104 ontrol s:
60-year neutron fluence for E>1 MeV is 4.6x1018 n/cm2 12 inches from RCS pipe OD 3D-38 Revision 1
Inside Containment Conditions/Parameter Abnormal Extreme Duration Notes up 1 (150°F) Abnormal Events emperature 150°F 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Note 1 ressure Atmospheric umidity 100% 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Note 1 adiation Same as normal hemistry None ubmergence None up 2 (250°F) Abnormal Events emperature 250°F 30 days Note 1 ressure 15 psig 30 days Note 1 umidity 100% 30 days Note 1 adiation Note 2 hemistry None ubmergence None s:
Parameter value is not maximum for full duration.
Minor increase over normal radiation conditions expected.
3D-39 Revision 1
Outside Containment Conditions/Parameter Abnormal Extreme Duration Notes ne 2 - Loss of AC Power Temperature Figure 3D.5-1 (Sheet 2) 7 days Note 3 Pressure Atmospheric Humidity 40 - 95% Note 2 Radiation Same as normal Chemistry/Submergence None ne 3 - Loss of HVAC Temperature Figure 3D.5-1 (Sheet 1) 7 days Pressure Atmospheric Note 1 Humidity 60 - 95% Note 2 Radiation Same as normal Chemistry/Submergence None ne 4 - Loss of AC Power Temperature 120°F max 10x4 hrs Pressure Atmospheric Humidity Same as normal Radiation Same as normal Chemistry/Submergence None ne 5 - Loss of AC Power Temperature 150°F max 10x4 hrs Pressure Atmospheric Humidity Same as normal Radiation Same as normal Chemistry/Submergence None ne 6 - Loss of AC Power Temperature 140°F max 10x4 hrs Pressure Atmospheric Humidity Same as normal Radiation Same as normal Chemistry/Submergence None 3D-40 Revision 1
Outside Containment Conditions/Parameter Abnormal Extreme Duration Notes 7 - Loss of AC Power mperature 114°F max 10x4 hrs essure Atmospheric umidity Same as normal adiation Same as normal hemistry/Submergence None s 8, 9, 10 mperature Same as normal essure Same as normal umidity Same as normal adiation Same as normal hemistry/Submergence None 11 - Loss of AC Power (Fuel Handling Area) mperature 212°F max 7 days essure Atmospheric Note 4 umidity 100%
adiation Same as normal hemistry/Submergence None s:
Main control room air pressure is maintained above a nominal value of atmospheric during accident conditions to prevent radioactive contaminant entry.
Figure 3D.5-1 Sheets 1 and 2 have two curves post-72 hours. The high curve represents the introduction of outside air that is high temperature, low humidity. The low curve represents the introduction of outside air that is low temperature, high humidity.
The EQ Programs will include both of these extremes.
Test environments resulting from rooms with equipment supplied by 24- and 72-hour batteries are shown on Sheet 2 for the dc equipment rooms 12203 and 12207 and for the I&C rooms 12302 and 12304. The 24-hour battery is disconnected at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The 72-hour battery is not disconnected. Environments resulting from rooms with equipment supplied by 24-hour batteries only, - that is, dc equipment rooms 12201 and 12205 and I&C rooms 12301 and 12305 - are enveloped by the environments shown on Sheet 2.
A relief panel is designed to open when the fuel handling area temperature exceeds 165°F.
3D-41 Revision 1
(See Table 3D.5-1 for environmental zones) 1 - Inside Containment Temperature and pressure See Figure 3D.5-8.
Submergence as applicable up to elevation 110-6 Radiation See Figures 3D.5-2 through 3D.5-5.
s 2, 3, 4, 6, 7, 8, 9, 11 (Same as abnormal - see Table 3D.5-4.)
s 5 and 10 - Outside Containment MSIV Compartments Temperature See Figure 3D.5-9.
Radiation See Figures 3D.5-4 and 3D.5-5.
3D-42 Revision 1
Environmental Qualification Component Material Property kets Compression set/elongation ngs Compression set/elongation phragms Elongation/tensile strength phragm support sheets Tensile strength/elongation ricant Viscosity/penetration m gear Flexural strength 3D-43 Revision 1
Figure 3D.5-1 (Sheet 1 of 3)
Typical Abnormal Environmental Test Profile: Main Control Room 3D-44 Revision 1
I&C and DC Equipment Rooms 130 120 110 Temperature (F)
Time: 0 to 72 hrs Time: 72 to 168 hrs 100 Relative Humidty: 40% Relative Humidity: 40%
90 80 70 Time: 72 to 168 hrs Time: -12 to 0 hrs Relative Humidity: 60% Relative Humidity: 95%
60
-12 12 36 60 84 108 132 156 Time (hr)
Figure 3D.5-1 (Sheet 2 of 3)
Typical Abnormal Environmental Test Profile: I&C and DC Equipment Rooms 3D-45 Revision 1
Figure 3D.5-1 (Sheet 3 of 3)
Typical Abnormal Environmental Test Profile: Voltage and Frequency Variations 3D-46 Revision 1
Figure 3D.5-2 Gamma Dose and Dose Rate Inside Containment After a LOCA 3D-47 Revision 1
Figure 3D.5-3 Beta Dose and Dose Rate Inside Containment After a LOCA 3D-48 Revision 1
Figure 3D.5-4 Gamma Dose and Dose Rate Inside Containment After a Steam Line Break 3D-49 Revision 1
Figure 3D.5-5 Beta Dose and Dose Rate Inside Containment After a Steam Line Break 3D-50 Revision 1
Figure 3D.5-7 Not Used.
3D-51 Revision 1
Figure 3D.5-8 (Sheet 1 of 2)
Typical Combined LOCA/SLB/FLB Inside Containment Temperature 3D-52 Revision 1
Figure 3D.5-8 (Sheet 2 of 2)
Typical Combined LOCA/SLB/FLB Inside Containment Pressure 3D-53 Revision 1
Figure 3D.5-9 (Sheet 1 of 2)
MSIV Compartment Response to MSLB (Short Term) 3D-54 Revision 1
Figure 3D.5-9 (Sheet 2 of 2)
MSIV Compartment Response to MSLB (Long Term) 3D-55 Revision 1
3D-56 Revision 1 equipment qualification data package consists of the following elements:
tion 1.0-Specifications tion 2.0-Qualification Program tion 3.0-Qualification by Test tion 4.0-Qualification by Analysis tion 5.0-Qualification by Experience tion 6.0-Qualification Program Conclusions le 1-Qualification Summary 3D-57 Revision 1
EQDP-_______
Rev. _______
{date issued}
EQUIPMENT QUALIFICATION DATA PACKAGE Equipment __________________________________________
Manufacturer __________________________________________
Model __________________________________________
Application __________________________________________
Environment: ____ Harsh ____ Mild Prepared by: __________________________________________
{name}
Reviewed by: __________________________________________
{name}
Approved by: __________________________________________
{name}
This document provides or summarizes the seismic and environmental qualification of the equipment identified above in accordance with the AP1000 EQ Program Methodology.
3D-58 Revision 1
1.0 SPECIFICATIONS 1.1 EQUIPMENT IDENTIFICATION: {create table(s) for details if a model series is to be qualified.}
Manufacturer ___________________________________
Model ___________________________________
Technical Manual ___________________________________
Drawings ___________________________________
Specification No. ___________________________________
Modifications ___________________________________
1.2 INSTALLATION REQUIREMENTS: {Cite vendor technical manual; details of mounting used for seismic test specimen(s);
include any special requirements unique to Class 1E service}
1.3 ELECTRICAL REQUIREMENTS
1.3.1 Voltage
1.3.2 Frequency
__________ {if powered by AC}
1.3.3 Load: __________ {as applicable}
1.3.4 Other
__________ {identify and address as needed}
1.4 AUXILIARY DEVICES: {These are devices required to be interfaced with the subject equipment to provide qualification or operability but not specifically included or addressed in this document.}
1.5 PREVENTATIVE MAINTENANCE: {Identify manufacturer recommended maintenance activities required as part of the qualification program. Identify activities that are required to support qualification or the qualified life. "None" shall mean that maintenance is not essential to qualification or the qualified life. The following statement may be used in cases where qualification is not contingent upon maintenance or surveillance activities:
"No preventive maintenance is required to support the equipment qualified life. This does not preclude development of a preventive maintenance program designed to enhance equipment performance and identify unanticipated equipment degradation as long as this program does not compromise the qualification status of the equipment. Surveillance activities may also be considered to support the basis for, and a possible extension, of the qualified life."}
1.6 SAFETY FUNCTIONS
{Specify known safety functions for which qualification is intended to apply.}
1.7 PERFORMANCE REQUIREMENTS(a) for: {RCS Loop RTDs}
Containment DBE(b) Conditions Normal Abnormal Test Parameter Conditions Conditions Abnormal Seismic LOCA 1.7.1 Time requirement 1.7.2 Performance 1.8 ENVIRONMENTAL CONDITIONS(a) for Same Function 1.8.1 Temperature (°F) 1.8.2 Pressure (psig) 1.8.3 Humidity (%RH) 1.8.4 Radiation (Rads) 3D-59 Revision 1
1.8.5 Chemicals 1.8.6 Vibration 1.8.7 Acceleration (g)
Notes: a: Test margin is not included in the parameters of this section.
b: DBE is the Design Basis Event.
{If more than one set of performance requirements and/or associated environmental conditions are to be specified, replicate these sections in pairs as "1.8 Performance ..." and "1.9 Environment ...", etc.}
3D-60 Revision 1
2.0 QUALIFICATION PROGRAM 2.1 PROGRAM OBJECTIVE The objective of this qualification program is to demonstrate, employing the recommended practices of Regulatory Guides 1.89 and 1.100 and IEEE 323-1974, 344-1987, {cite others as applicable} capability of the {Equipment description} to perform its/their safety related function(s) described in EQDP Section 1.7 while exposed to the applicable conditions and events defined in EQDP Section 1.8.
{Narrative should introduce an outline of the program plan. Table below to be completed as graphic reference. Table shall not be abbreviated; items must appear and be addressed by direct response.}
2.2 REFERENCES
{List test report(s) and information sources cited in this document}
3D-61 Revision 1
Qualification Method(s)
CONDITION TEST ANALYSIS OTHER Aging:
Thermal Radiation Vibrational Operational Cycling Electrical Mechanical Abnormal Environment Inadvertent ADS Actuation Seismic LOCA HELB Inside Containment HELB Outside Containment Post-accident Aging NOTES:
{All spaces above to be noted as "Yes," "No," or "Note #." Notes will be appended to the table. Notes will also include items "Not Applicable" with terse explanation and/or forwarding reference.}
3D-62 Revision 1
3.0 QUALIFICATION BY TEST (TEST PLAN AND
SUMMARY
)
3.1 SPECIMEN DESCRIPTION
{Identify the item or items to be tested}
3.2 NUMBER TESTED
{If more than one type is to be tested, identify how many of each. Subsequent Sections should clarify specifics for each.}
3.3 MOUNTING
{Identify specific seismic mounting details, referencing applicable drawings, instructions, documents. Note existence of differences from manufacturer recommendations}
3.4 CONNECTIONS
{Identify interfaces, both electrical and mechanical, identify any connectors or sealing assemblies used which are not provided with the equipment, or are not covered by this qualification.
3.5 TEST SEQUENCE PREFERRED This section identifies the preferred test sequences as specified in IEEE 323-1974.
3.5.1 Inspection of Test Item 3.5.2 Operation (Normal Condition) 3.5.3 Operation (Performance Specifications Extremes: Section 1) 3.5.4 Simulated Aging 3.5.5 Vibration/Seismic 3.5.6 Operation (Simulated High Energy Line Break Conditions) 3.5.7 Operation (Simulated Post-HELB Conditions) 3.5.8 Inspection 3.6 TEST SEQUENCE ACTUAL This section identifies the actual test sequence which constitutes the qualification program for this equipment. A justification for anything other than the preferred sequence is provided.
Test Sequence (from Section 3.5):
{List and explain; provide forwarding references to subsequent subsections as necessary}
3D-63 Revision 1
3.7 SERVICE CONDITIONS TO BE SIMULATED BY TEST(1)
Normal Abnormal Seismic HELB Post-HELB 3.7.1 Temperature (°F) 3.7.2 Pressure (psig) 3.7.3 Humidity (% RH) 3.7.4 Radiation (Rads) 3.7.5 Chemicals 3.7.6 Vibration 3.7.7 Seismic (g)
(1) Test parameter margins are included for the worst-case known requirements applicable to the equipment type. Margin for a specific parameter is dependent on the requirements of each application or location for the equipment; these may vary.
(2) Post-accident operability addressed through simulated thermal aging. Temperature and other parameters are selected to envelop the requirements.
3D-64 Revision 1
3.8 MEASURED VARIABLES This section tabulates the variables and parameters required to be measured during each of the following tests in the qualification test sequence.
Tests: {example}
A: Thermal Aging B: Mechanical Cycling C: Irradiation D: Seismic Test E: HELB Test 3.8.1 Category I - Environment Required Not Required 3.8.1.1 Temperature ... ...
3.8.1.2 Pressure ... ...
3.8.1.3 Moisture ... ...
3.8.1.4 Gas Composition ... ...
3.8.1.5 Vibration ... ...
3.8.1.6 Time ... ...
3.8.2 Category II - Input Electrical Characteristics 3.8.2.1 Voltage ... ...
3.8.2.2 Current ... ...
3.8.2.3 Frequency ... ...
3.8.2.4 Power ... ...
3.8.2.5 Other ... ...
3.8.3 Category III - Fluid Characteristics 3.8.3.1 Chemical Composition ... ...
3.8.3.2 Flowrate ... ...
3.8.3.3 Spray ... ...
3.8.3.4 Temperature ... ...
3.8.4 Category IV - Radiological Features 3.8.4.1 Energy Type ... ...
3.8.4.2 Energy Level ... ...
3.8.4.3 Dose Rate ... ...
3.8.4.4 Integrated Dose ... ...
3.8.5 Category V - Electrical Characteristics 3.8.5.1 Insulation Resistance ... ...
3.8.5.2 Output Voltage ... ...
3.8.5.3 Output Current ... ...
3.8.5.4 Output Power ... ...
3.8.5.5 Response Time ... ...
3.8.5.6 Frequency Characteristics ... ...
3.8.5.7 Simulated Load ... ...
3D-65 Revision 1
Required Not Required 3.8.6 Category VI - Mechanical Characteristics 3.8.6.1 Thrust ... ...
3.8.6.2 Torque ... ...
3.8.6.3 Time ... ...
3.8.6.4 Load Profile ... ...
3.8.7 Category VII - Auxiliary Equipment 3.8.7.1 {as applicable, also see ... ...
Section 1.4 of EQDP}
3D-66 Revision 1
3.9 TYPE TEST
SUMMARY
3.9.1 Normal Environment Testing Operation of the {equipment} under normal conditions is demonstrated by {discuss test, checks, et. al. which provide baseline performance data} ... , as reported in Reference .
3.9.2 Abnormal Environment Testing Operation of the {equipment} under abnormal conditions is demonstrated by {discuss test, checks, et. al. which provide baseline performance data} ... , as reported in Reference .
3.9.3 Aging Simulation Procedure
{Describe the aging mechanisms simulated and the sequence, including justifications as necessary.}
The test units were pre-conditioned to simulate an aged condition prior to subjecting them to the Design Basis Event (DBE) seismic and environmental conditions/simulation. The aged condition was achieved by separate phases of {accelerated thermal aging, thermal cycling, and radiation exposure to a total integrated gamma dose equivalent to a twenty-year normal dose plus the design basis accident dose, and accelerated flow induced and pipe vibration simulation}. Through all the pre-conditioning phases, the {equipment, performance} were monitored to verify {continuous operation}.
3.9.3.1 Design Life: {Also, justification of the bases for a design life goal should be provided, when used in mild-environment programs. Generally inapplicable to harsh-environment programs.}
3.9.3.2 Shelf Life: {Though not typically applicable, state any limitation in life, as well as conditions which may be detrimental if known.}
3.9.3.3 Thermal Aging: The qualified life is years based on an ambient temperature of { °C ( °F) and a °C temperature rise due to }. Calculations are based on a test temperature of , test duration of hours, and an activation of eV (See References x, et al.)}.
3.9.3.4 Radiation Aging: The qualified life is limited by the expected radiation during the -year life and the Design Basis Event. {Subtract accident TID from qualified TID; account for margin, remainder is to be compared to normal/abnormal radiation requirements to yield life limits.}
3.9.3.5 Operating Cycles: {Expected number of electrical and/or mechanical cycles, or numbers of actuations, as applicable.
Estimated on the basis of the expected for the design, qualified, installed life of the equipment. Specification may be on a per annum or a per fuel cycle basis. Compare to cycle life data from test.}
3.9.3.6 Vibration Aging: {present bases; refer to test profile and/or Subsection 3.9.4}.
3.9.4 Seismic Tests The seismic testing reported in Reference x was completed on aged equipment employing {method(s)} in accordance with Regulatory Guide 1.100 and IEEE 344-1987. ... {Summarize equipment condition and/or performance versus the acceptance criteria.} ... Actual margin should be determined for each application/location throughout the plant and verified to meet or exceed the margin requirements.
{Discuss or reference discussion of test anomalies.}
3D-67 Revision 1
3.9.5 High Energy Line Break/Post HELB Simulation The {equipment} were subjected to the HELB simulation temperature/pressure profile of Figure x. Following the °F temperature peak, the temperature gradually declines to °F and is held at saturated steam conditions for days, simulating a period of Post-HELB operation. The test data and activation energy specified in Subsection 3.9.3.3 can be used to determine margin in post-accident aging for each application/location of the equipment.
{Summarize equipment condition and/or performance versus the criteria}
{Discuss or reference discussion of test anomalies.}
3D-68 Revision 1
4.0 QUALIFICATION BY ANALYSIS The AP1000 EQ Program does permit qualification solely on the basis of analyses for equipment outside the scope of 10CFR50.49.
The following subsections discuss each of the analyses preformed, its test basis and justification, and summarizes conclusions documented in References x; et. al., which provided detailed accounts of each analysis.
{Each subsection will address a particular analysis, if more than one is performed to support qualification.}
4.x (EXAMPLE)
{The purpose and objective will be identified here. Subsections will provide necessary details per the following format.}
4.x.1 {Equipment, Characteristic or Aspect} Analyzed
{A general description of the equipment and its function based on applicable equipment and mounting drawings, and purchase orders.}
4.x.2 Equipment Specification(s)
{The applicable design standards shall be documented including any limitations imposed by the equipment specification. Installation detail considered or represented are to be included.}
4.x.3 Methods and Codes
{Description of analytical methods or techniques, computer program, mathematical model(s) used, and the method(s) of verification}
4.x.4 Acceptance Criteria
{The specific safety function(s), postulated failure modes, or the failure effects to be demonstrated by analysis.}
4.x.5 Model
{Description of mathematical model of equipment or feature analyzed.}
4.x.6 Assumptions and Justifications
{EXAMPLES: Description of the loading conditions to be used. Summary of stresses to be considered.}
4.x.7 Impact to Safety Function
{Summarize analytically established performance characteristics and their acceptability. Discussion and summary of the analytical results which demonstrate equipment structural integrity and, where appropriate, operability. Particular to cabinets, critical deflections should be determined and included in mounting requirements for spacing with respect to other equipment and structures.}
4.x.8 Conclusions
{Descriptive summary, including any conditions imposed on qualification or use; qualified life, limitations, surveillance/maintenance requirements, et. al.} Further discussion of this analysis is presented in Reference x.
4.Y ENVIRONMENTAL QUALIFICATION ANALYSIS FOR {VALVE SOFT PARTS}
{purpose and objective}
4.Y.1 Equipment Identification
{Per Subsection 6.2.3.1}
3D-69 Revision 1
4.Y.2 Component Identification
{Per Subsection 6.2.3.1}
4.Y.3 Safety Related Functions
{Per Subsection 6.2.3.2}
4.Y.4 Component Acceptance Criteria
{Per Subsection 6.2.3.3}
4.Y.5 Service Conditions
{Per Subsection 6.2.3.4}
4.Y.6 Potential Failure Modes
{Per Subsection 6.2.3.5}
4.Y.7 Identify the Environmental Effects on Material Properties Each non-metallic, including lubricants, is evaluated to determine the effect of the environmental conditions on the material properties.
For each non-metallic, a radiation threshold level and maximum service temperature is identified.
The radiation threshold level and the maximum service temperature are identified using materials handbooks, textbooks, government and industry reports, and laboratory data. If the evaluation indicates that the lowest levels may be exceeded for certain equipment, higher levels are identified at which varying degrees of material degradation may occur.
Mechanical equipment is highly resistive to degradation due to elevated humidity levels: therefore, relative humidity is not included as a parameter to be evaluated for environmental qualification. Pressure can be discounted for most equipment types, as there are no foreseen failures due to elevated pressure levels for most mechanical equipment. However, pressure must be addressed in the evaluation.
The susceptibility of the non-metallic material to the chemicals due to the design basis accident and exposure to the process fluid is evaluated. The material information in the chemical handbooks is an acceptable source of qualification documentation.
4.Y.7.1 Perform Thermal Aging Analysis Aging analysis is performed for organic materials. Mineral-based subcomponents are not considered to be sensitive to thermal aging during the design life of a plant and, therefore, are not analyzed.
Aging in mechanical components is associated with corrosion, erosion, particle deposits and embrittlement. In new construction, corrosion and erosion are considered by providing additional material thickness as a corrosion or erosion allowance above the required design. The other aging phenomena are considered during inservice inspections of operating components in accordance with ASME Code,Section XI. Aging qualification of metallic parts of equipment except for corrosion and erosion is in compliance with ASME Code,Section XI, therefore aging effects on metallic components are not addressed herein.
The non-metallic material analysis for determining the expected qualified thermal life is performed using Arrhenius methodology. The thermal input during the operating time, as explained below, is deducted from the tested thermal aging of the material at service temperature to obtain the qualified life.
The component is evaluated for the specified post-accident operating time. The thermal input from the postulated accident profile (i.e., LOCA/MSLB) for the duration of the specified operating time is compared to the material thermal aging data. The Arrhenius model is used to perform this comparison. The component is evaluated for the maximum post-accident operating time unless a system analysis is performed to justify shorter operating times.
3D-70 Revision 1
Analysis of the non-metallics should also take into account any degradation of the part due to its use in dynamic modes (i.e., moving part).
4.Y.7.2 Evaluate the Environmental Effects on Equipment Safety-Related Function A conservative initial screening of the non-metallic subcomponents is made by comparison of the material capabilities (threshold radiation level and maximum service temperature) with the maximum postulated environmental conditions. If the threshold radiation values and the maximum service temperatures are above the maximum postulated environmental conditions, and if the material aging analysis demonstrates a service life sufficient to survive the accident duration, then the material is considered acceptable.
Those items which are not shown to be acceptable based on the above comparison are evaluated in further detail regarding:
- extent of material degradation
- material properties affected
- equipment/subcomponent function
- extent of equipment functional degradation
- location-specific environmental conditions 4.Y.8 Conclusions
{Per subsection 3D.6.2.3.7}
4.Y.9 EQ Maintenance Requirements
{Per subsection 3D.6.2.3.8}
3D-71 Revision 1
5.0 QUALIFICATION BY EXPERIENCE This method of qualification is not used.
3D-72 Revision 1
6.0 QUALIFICATION PROGRAM CONCLUSIONS 6.1 AGING
{Discuss specifics and state on limitations or requirements; specifics with respect to:
- Design Life Goal
- Thermal Aging
- Radiation Aging
- Operating Cycles
- Vibration Aging}
6.2 DBE QUALIFICATIONS 6.3 PROGRAM CONCLUSIONS The qualification of the {equipment} is demonstrated by the completion of the simulated aging and Design Basis Event testing described herein and reported in Reference {1}.
{State any conditions imposed on qualification or qualified life, cite any lessons learned which necessitate future user actions to preserve continued qualification}
{Refer to Table 1}
3D-73 Revision 1
Table 1 QUALIFICATION
SUMMARY
SYSTEM {RPS}
CATEGORY Category(1) {a}
LOCATION {Containment bldg.}
STRUCTURE/AREA {Zone Number}
EQUIPMENT TYPE {pressure transmitter }
MANUFACTURER { }
MODEL { }
QUAL ENVIRONMENTAL EXTREMES PARAMETER METHOD(2) QUALIFIED(3) SPECIFIED(4) NOTES NORMAL ABNORMAL QUALIFIED LIFE {5}
SEISMIC {Both} Figure x {Ref; Fig.}
ACCIDENT Figure x {Ref; Fig.}
Temperature {Test} °F Pressure {Test} psig Rel. humidity {Test} %
Radiation {Both} E+06 R()
{Both} E+06 R()
Chemistry {Test} {Note 6}
Operability {Both}
Accuracy {Test}
NOTES:
- 1. Equipment category as per NUREG-0588, Appendix E, Section 2.
- 2. Qual. Methods are: Test, Analysis, Both (Test & Anal.), or Other.
- 3. Qualified values are test extremes which include margin.
- 4. Environmental parameters for the plant location are to be inserted. If more than one applicable, most extreme are to be cited
- 5. Qualified life estimated on basis of maximum normal temperature of °C ( °F) and a temperature rise of °C
( °F).
- 6. Chemistry Conditions: {pH and composition}.
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Introduction tated in IEEE 323, aging of Class 1E equipment during normal service is considered as an gral part of the qualification program. The objective is not to address random age-induced failures occur in-service and are detected by periodic testing and maintenance programs. The objective address the concern that some aging mechanisms, when considered in conjunction with the cified design basis events (DBE), may have the potential for common mode failure.
AP1000 equipment qualification program addresses the aging concern and makes maximum of available data and experience on aging mechanisms. This approach places primary emphasis ommon mode failures due to enveloping design basis events. For example, reasonable urance against common mode failures being induced because of a loss of heating, ventilation, air conditioning (HVAC) is provided by adequate design, normal maintenance, and calibration edures.
Objectives objectives of the aging evaluation program follow:
To establish, where possible, the effects of the degradation due to aging mechanisms that occur before the occurrence of an accident, when safety-related equipment is called upon to function To provide increased confidence that safety-related equipment performs its safety-related function under the specified service condition.
Basic Approach general approach to addressing aging allocates equipment to one of two subprograms (A or B).
Subprogram A includes electrical equipment required to perform a safety-related function in a high-energy line break (HELB) environment. For this equipment an aging simulation is included as part of the equipment qualification test sequence. The equipment is energized during the aging simulation.
Subprogram B includes equipment required to mitigate high-energy line breaks but which, due to its location, is isolated from any adverse external environment resulting from the accident. For equipment in Subprogram B the single design basis event capable of producing an adverse environment at the equipment location is the seismic event. Aging, for Subprogram B, is not included in the equipment qualification test sequence. Significant aging mechanisms are determined by evaluation of available test data. Generally, this data is from separate programs conducted to demonstrate that aged components continue to meet manufacturer's performance specifications under applicable seismic design basis event conditions and that seismic testing of unaged equipment is not invalidated by anticipated aging mechanisms.
Subprogram A trical equipment required to perform a safety-related function in a high-energy line break (such loss of coolant accident, feed line break, or steam line break) environment is included in 3D-75 Revision 1
1 Scope typical equipment scope and aging mechanisms applied under Subprogram A are shown in les 3D.B-1 and 3D.B-2, respectively. The equipment selected is that Class 1E equipment lified to operate in a high-energy line break environment. The aging mechanisms discussed next those to which the equipment may be potentially sensitive in its installed location.
2 Aging Mechanisms aging mechanisms that could potentially affect electrical equipment in Subprogram A are ussed under the following headings:
e, in conjunction with:
Operational stresses (current, voltage, operating cycles, Joulean self-heating)
(External stresses (thermal, vibration, radiation, humidity, seismic).
aging mechanisms considered potentially significant and to be simulated are identified in le 3D.B-2 for each item of equipment in Subprogram A. Where applied, the aging mechanisms simulated as described in the following discussions.
3 Time equipment subject to high-energy line break conditions, the most significant in-service aging hanisms (that is, radiation and thermal) come into effect during reactor operation. Consequently, n be assumed that the "aging clock" starts on plant startup.
4 Operational Stresses ctrical Cycling trical supplies to safety-related equipment are, in general, highly stable. So aging effects due to ply cycling during service are not anticipated. Where the equipment is anticipated to experience tiple startup and shutdown cycles, the equipment is electrically cycled to simulate the number of cipated startup and shutdown cycles plus 10 percent.
hanical Cycling ng effects resulting from anticipated mechanical cycling of the equipment are simulated by lying, as a minimum, the number of cycles estimated to occur during the target qualified life plus ercent. Mechanical cycling covers such operations as switching and relay actuation.
lean Self-Heating ere the equipment is not aged in a live condition, the aging effects resulting from Joulean self-ting are recognized by employing the equipment operating temperature as the datum perature for assessing the accelerated thermal aging parameters to be employed.
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rmal effects are considered one of the most significant aging mechanisms to address. The ipment is thermally aged to simulate an end-of-qualified-life condition using the Arrhenius model stablish the appropriate conditioning period at elevated temperature. Where data is not available stablish the model parameters for the materials employed, a verifiably conservative value of eV is used for activation energy (Attachment D).
each piece of equipment an appropriate normal and abnormal operating temperature and an ociated time history are determined for inclusion in the Arrhenius model. The equipment perature is determined by the addition of an appropriate equipment specific T to the external ient temperature. Attachment D also provides information concerning the determination of ropriate ambient temperatures and time-temperature histories for use in thermal aging evaluation quipment. Post-accident thermal aging is included by recognizing the higher post-accident ient temperatures in determining the parameters employed for the post-accident accelerated mal aging simulation.
ervice Vibration majority of safety-related electrical equipment has a proven history of in-plant service. Thus, it is kely that a significant, undetected, failure mechanism exists because of low-level, in-plant ation. In addition, a simulation of earthquakes smaller than the safe shutdown earthquake (SSE) loyed during equipment and component seismic testing give added confidence that this potential g mechanism is covered (See Attachment E, Section E.4.4). For line-mounted equipment, ervice pipe and flow induced vibration may be significant. As a consequence, an additional ation aging step is included in the aging sequence as indicated for certain items of equipment in le 3D.B-2. (See Attachment E, Subsection E.5.2.4.)
iation iation during normal operation is not considered an aging mechanism for equipment subject to ervice integrated doses less than 104 rads. Research has established that no aging mechanisms measurable below 104 rads (Attachment C) for materials and most components supplied in ty-related electrical equipment. Some devices may have performance limitations below 104 rads.
radiation doses in excess of 104 rads, the equipment is irradiated using a gamma () source to a e equivalent to the estimated dose to be incurred during normal operation for the target qualified The estimated doses employed are specified in the equipment qualification data package, section 1.8.4, and are based on a 100 percent load factor, including appropriate margin. For program A equipment, the equivalent accident dose is usually applied before design basis event ing.
midity use of materials significantly affected by humidity is avoided. For equipment subject to high rgy line break environments, the aging effects due to humidity during normal operation are judged e insignificant compared to the effects of the high-temperature steam accident simulation.
refore, no additional humidity aging simulation is required.
smic Aging potential aging effects of low-level seismic activity and some low-level, in-plant vibration are ressed by employing a simulation of five earthquakes of 50 percent of the magnitude of a safe tdown earthquake before seismic testing of the aged equipment.
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ss environments are applied simultaneously. The potential for significant synergistic effects is ressed by the conservatisms inherent in using the "worst-case" aging sequence, conservative elerated aging parameters and conservative, design basis event test levels which provide fidence that any synergistic effects are enveloped.
7 Design Basis Event Testing ign basis event testing subsequent to equipment aging is discussed in Appendix 3D as to elines for defining high-energy line break environments and seismic conditions. Testing for ipment specific test environments and seismic parameters is discussed in Attachment A, tion 3.0.
8 Aging Sequence aging mechanisms applied to equipment subject to high-energy line break environments are rmined by definition of the aging environments at the equipment location and by a subsequent luation of the sensitivity of the equipment to these environments. If the sensitivity of the ipment is not known, aging mechanisms are simulated by conservative methods as previously cribed. Those aging mechanisms that are simulated for typical equipment subject to high-energy break environments are shown in Table 3D.B-2.
order in which each of the aging mechanisms is applied is as shown in Table 3D.B-2. This order onsidered to be conservative, as no aging mechanism is anticipated to be capable of reducing the act of the previously applied mechanisms. As an example, thermal aging is applied before ation aging to preclude the annealing out of radiation-induced defects. Similarly, the effects of hanical aging are considered more significant when applied to equipment that has already been aged to address thermal and radiation phenomena.
9 Performance Criterion basic acceptance criterion is that the qualification tests demonstrate the capability of the aged ipment to perform prespecified, safety-related functions consistent with meeting the performance cification of Attachment A, Section 1.7 of the applicable equipment qualification data packages e exposed to the associated environmental conditions defined in Attachment A, Section 1.8.
10 Failure Treatment en thermal aging is simulated at an equipment level, a conservative value for the activation rgy is assumed for the components composing the equipment. As a consequence, many ponents are grossly overaged, and failure of some of the components is expected during the g simulation. When three test units are preaged, in the event of such failure(s), one of the wing options is selected.
when a particular component fails in one of the three test units, the failure is considered random. The failed component is replaced by a new component, and the test is continued when a particular component fails in more than one of the three test units, either:
- 1. the failed components are replaced by new identical components and the aging simulation continued. The claimed qualified life of the unit is consistent with the minimum aging period simulated by at least two of the three units; or 3D-78 Revision 1
- 3. the failed components are replaced by a different type of component which is aged for a period equal to the test units.
en less than three test samples prevent such a conclusion from being reached, any failures are stigated to ascertain whether the failure mechanism is of common mode origin. Should a mon mode failure mechanism be identified as having caused the failure, a design change is lemented to eliminate the problem. Supplemental or repeat tests will be completed to onstrate compliance with the acceptance criteria.
Subprogram B program B includes Class 1E equipment not required to perform a safety related function in a
-energy line break environment. It involves a review of available information to demonstrate the ence of significant in-service aging mechanisms. For equipment allocated to this subprogram, the le design basis event capable of producing an adverse environment at the equipment location is seismic event. Seismic testing completed on unaged equipment is verified as valid by onstrating via this subprogram that no available information suggests that aged materials and ponents would not continue to meet their design specification during a seismic event.
1 Scope program B includes both a review of material analysis and the results of a component testing gram for equipment not required to perform a safety-related function in a high-energy line break ironment. Equipment is included that is required to mitigate high-energy line breaks but which, ause of the equipment location, is isolated from the adverse environment resulting from the dent. Typical equipment allocated to Subprogram B is identified in Table 3D.B-1.
2 Performance Criteria ilable Material Analysis - For equipment and components for which aging is addressed by luation of appropriate mechanisms, the basic performance criterion is that the evaluation of test demonstrates the effect of aging is minor and does not affect the capability of the aged ipment to perform prespecified functions. This is consistent with meeting the performance cification of Attachment A, Section 1.7 of the applicable equipment qualification data package e exposed to the associated environmental conditions defined in Attachment A, Section 1.8.
ilable Component Aging Data - Random component failure or unacceptable performance due to g is detected by routine maintenance and equipment calibration during service. The objective of program B is to provide reasonable assurance that a seismic event does not constitute a mon mode failure mechanism capable of inducing unacceptable performance characteristics in d components. Consequently, the single performance criterion for the aging portion of the lification sequence requires that the component not fail to perform its general function, not that component meets the original design and procurement specifications.
the seismic event simulation, the component is considered acceptable if, during and after the ulation, it does not exhibit any temporary or permanent step change in performance racteristics. Failure of one of three components tested is considered a random failure, subject to nvestigation concluding the observed failure is not common mode.
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lution of qualification with respect to age:
Establish a maintenance and surveillance program Replace the materials or components with those constructed of materials of known acceptable characteristics.
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Aging Method Equipment program A Valve Motor Operators Solenoid Valves Externally Mounted Limit Switches Pressure Transmitter (Group A)
Differential Pressure Transmitter (Group A)
Resistance Temperature Detectors Neutron Detectors Pressure Sensor Batteries*
program B Pressure Transmitter (Group B)
Differential Pressure Transmitter (Group B)
Main Control Board Switch Modules Recorders (Post-Accident Monitoring)
Indicators (Post-Accident Monitoring)
Instrument Bus Distribution Panels Instrument Bus Power Supply (Static Inverter)
Motor Control Centers Integrated Protection Cabinets (IPC)
Engineered Safety Features Actuation Cabinets (ESFAC)
Logic Cabinets Reactor Trip Switchgear Reactor Coolant Pump Switchgear To comply with R.G. 1.158 3D-81 Revision 1
Aging Mechanisms DBE Equipment Location Subprogram Burn-in Thermal Radiation Mechanical Vibration Electrical Seismic Seismic HELB ty-related Valve Motor I/C A X X X X X X X ators O/C A X X X X X X y-related Solenoid I/C A X X X X X X X s O/C A X X X X X X y-related Externally I/C A X X X X X X X nted Limit Switches O/C A X X X X X X sure Transmitters I/C&OC A X X X X X X rential Pressure I/C&OC A X X X X X X smitters stance Temperature I/C A X X X X X X ctors: Well Mounted re Neutron Detectors I/C A X X X X X sure Sensor I/C A X X X 3D-82 Revision 1
OPERTIES OF MATERIALS Introduction potential common-mode failure mechanism to consider in the qualification of safety-related ipment is gamma radiation. As part of a qualification program, the effect of gamma radiation dose onsidered for two purposes: as a component of the high-energy line break environment and as a ntial aging mechanism that could reduce the capability of safety-related equipment to perform ty-related functions under design basis event conditions (seismic or high-energy line break).
scope of this attachment is limited to consideration of the effect of radiation for that substantial ion of equipment that does not experience an adverse change in external environment as a result high-energy line break, and for which, therefore, the only gamma radiation concern is an ervice aging mechanism.
attachment assumes that the equipment contains devices that have been selected for ormance through the total integrated dose expected in service. For example, devices such as grated circuits may have a limit of 1000 rads established, in which case the following discussion lies for its installed life. The information in this attachment is not adequate to be applied to ipment that must perform its function in a high-energy line break.
primary purpose of equipment qualification is to reduce the potential for common-cause failures to environmental effects during the qualified life. Random failures that inevitably occur inservice accommodated by the redundancy and diversity of the design of safety-related systems.
hermore, in-service maintenance and testing programs are designed to detect such random res. The chances of two identical components that perform identical functions failing during the e limited time period in between routine tests considered insignificant because of the following:
General low failure rate of components used in nuclear equipment Minor differences in component material or geometric tolerances or both Minor differences in operating environment.
refore, failures that are induced in components by normal background gamma radiation below rads (103 rads for some devices) alone are considered to be random. Thus, the only gamma ation concern addressed for equipment not subject to an adverse high-energy line break ironment is the potential for an aging mechanism resulting in a deterioration in component perties such that, when subject to seismic stress, a common-cause failure results. When sidering such a failure mode, the aging mechanism of concern is not one that affects the electrical perties of components but one that reduces the mechanical strength and flexibility of components.
Scope report summarizes available information concerning the effects of gamma radiation on material hanical properties. It justifies that for a gamma dose of less than 104 rads there are no ervable radiation effects that impact material mechanical properties. Of the materials stigated, only Teflon TFE is subject to an alteration of mechanical properties for a gamma dose of than 105 rads. Information is drawn from several sources listed as references in Section C.5.
y include various texts concerning radiation effects and damage and pertinent reports.
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e rates, which is of negligible significance here), and some displacement damage caused by
-energy photons. Some other types of radiation have effects similar to those induced by gamma ation. This allows the use of data obtained from exposure of material to an alternate radiation to ide limited information concerning the effects of exposure to gamma radiation.
example, the primary consequence of fast-neutron bombardment of material is atom lacement. Therefore, if the effect of radiation on a material property is primarily dependent on m displacement, it is inferred that for an equivalent dose (rads) of gamma and fast-neutron ation, data obtained from neutron irradiation provides a conservative estimate of the effect of ma irradiation in producing displacements.
same type of inference is drawn for the ionization effect of charged particle (for example, tron, proton, alpha particle) irradiation. Charged particles do not have the penetration capability gamma or neutron radiations exhibit as a result of extensive interaction between charged icles and atomic charge centers.
le 3D.C-1 summarizes information derived from the listed references. The information relates to effect of gamma radiation on material mechanical properties. Table 3D.C-1 presents either the shold dose (that dose at which an effect on any mechanical property can first be detected) or, the e that results in the identified effect. This provides a general indication of the susceptibility of erial mechanical properties to gamma radiation.
evaluation of the information available on inorganic materials summarized in Table 3D.C-1 shows the mechanical damage threshold for gamma radiation is many orders of magnitude greater than rads. For the organic materials listed in Table 3D.C-1, a histogram comparing threshold dose l and frequency of material susceptibility is provided. In instances for which a material threshold e is not indicated in Table 3D.C-1, a threshold value is assumed which is one order of magnitude er than the indicated damage dose. Where information is available, referenced documents cate that the difference between threshold dose and 25 percent damage dose is about a factor of
- e. Thus, a factor of 10 supplies substantial margin in estimating the threshold dose level.
re 3D.C-1 shows that any indications of mechanical property damage thresholds below 104 rads ld be extremely unusual.
references listed do not identify the existence of materials whose mechanical properties are riorated when exposed to a gamma radiation dose up to 104 rads. So it can be concluded that mon-cause failures do not occur in electrical equipment during or after a seismic event as a lt of radiation-induced degradation up to 104 rads.
is supported by NRC documentation available as an attachment to "Guidelines for Evaluating ironmental Qualification of Class 1E Electrical Equipment in Operating Reactors," which provides her justification for the use of 104 rads as a threshold for mechanical damage. The NRC rmation appears to be consistent with the information provided in Table 3D.C-1.
Conclusions Class 1E equipment subject to a lifetime gamma dose of up to 104 rads, it is not necessary to ress radiation aging for qualification purposes provided that the equipment is not required to orm a safety-related function in a high-energy line break environment.
previously noted, this appendix does not apply to electrical properties of components in safety-ted equipment.
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R. G. Krieger Publishing Co., 1986.
NASA Tech Brief Vol. 10, No. 5, Item #3, "Response of Dielectrics to Space Radiation,"
October 1986.
IRT Study 4331-006, "Design Guidelines for Transient Radiation Effects on Tactical Army Systems," Harry Diamond Labs, June 12, 1981.
Regulatory Guide 1.89, Rev. 1.
Hanks, C. L., and Hamman, D. J., "The Effect of Radiation on Electrical Insulating Materials," REIC Report No. 46, Radiation Effects Information Center, Battelle Memorial Institute, Columbus, Ohio, June 1969.
Kangilaski, M., "The Effects of Neutron Irradiation on Structural Materials," REIC Report No. 45, Radiation Effects Information Center, Battelle Memorial Institute, Columbus, Ohio, June 1967.
Hanks, C. L., and Hamman, D. J., "The Effect of Nuclear Radiation on Capacitors," REIC Report No. 44, Radiation Effects Information Center, Battelle Memorial Institute, Columbus, Ohio, December 1966.
Chapin, W. E., Drennan, J. E., and Hamman, D. J., "The Effect of Nuclear Radiation on Transducers," REIC Report No. 43, Radiation Effects Information Center, Battelle Memorial Institute, Columbus, Ohio, October 1966.
Drennan, J. E., and Hamman, D. J., "Space Radiation Damage to Electronic Components and Materials," REIC Report No. 39, Radiation Effects Information Center, Battelle Memorial Institute, Columbus, Ohio, January 1966.
Larin, F., "Radiation Effects in Semiconductor Devices," John Wiley and Sons, New York, 1968.
Billington, D. S., and Crawford, J. H., "Radiation Damage in Solids," Princeton University Press, Princeton, New Jersey, 1961.
Corbett, J. W., "Electron Radiation Damage in Semiconductors and Metals," Academic Press, New York, 1966.
Ricketts, L. W., "Fundamentals of Nuclear Hardening of Electronic Equipment,"
Wiley-Interscience, New York, 1972.
Kircher, J. F., and Bowman, Richard E., "Effect of Radiation on Materials and Components," Reinhold Publishing Corp., New York, 1964.
Bolt, R. O., and Carroll, J. G., "Radiation Effects on Organic Materials," Academic Press, New York, 1963.
Kaplan, Irvin, "Nuclear Physics," Addison-Wesley, 1962.
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of Material Mechanical Properties Material Mechanical Damage Threshold Dose for Comments tural Metals 1019 n/cm2 (fast neutron spectrum) Similar to cold work (1010 rads) anic Materials ~1017 n/cm2 (fast neutron spectrum) Borated materials have lower threshold values for neutron irradiation.
omers tural Rubber 2x106 rads(C) lyurethane Rubber 9x105 rads(C) yrene-Butadiene Rubber 2x106 rads(C) trile Rubber 7x106 rads(C) Compression set is 25% degraded oprene Rubber 7x106 rads(C) palon ~107 rads(C) Variable rylic Rubber 9x107 rads(C) Variable icone Rubber 107 rads(C) ~25% damage uorocarbon Rubber 9x107 rads(C) ~25% hardness, 80% elongation lysulfate Rubber 108 rads(C) tyl Rubber 107 rads(C) ~25% damage rad (C) is the field of radiation that will produce 100 ergs/gm in carbon.
ic flon TFE 1.7x104 rads(C) l-F 1.3x106 rads(C) lyethylene 107 rads(C) lystyrene 108 rads ylar 106 rads(C) Conservative lyamide (Nylon) 8.6x105 rads(C) allyl Phthalate 108 rads(C) lypropylene 107 rads(C) lyurethane 7x108 rads(C) 3D-86 Revision 1
Material Mechanical Damage Threshold Dose for Comments ic (Continued) nar (400) 107 rads(C) rylics 8.2x105 rads mino Resins 106 rads omatic Amide-Imide 107 rads sins 107 rads llulose Derivatives 3x107 rads 25% damage lyester, Glass Filled 8.7x108 rads enolics 3x108 rads(C) 25% damage icones 108 rads(C) lycarbonate Resins 5x107 rads 25% damage to elongation lyesters ~ 105 - 106 rads yrene Polymers 4x107 rads(C) yrene Copolymers 4x107 rads(C) 25% damage nyl Polymers 1.4x106 - 8.8x107 rads(C) nyl Copolymers 1.4x106 - 8.8x107 rads(C) psulating Compounds V 501 2x106 rads lgard 182 2x106 rads lgard 1383 2x106 rads lyurethane Foam 2x106 rads oxies 109 rads 3D-87 Revision 1
Figure 3D.C-1 Histogram of Threshold Gamma Dose for Mechanical Damage to Elastomers, Plastics, and Encapsulation Compounds 3D-88 Revision 1
Introduction chment B describes the approach employed in the AP1000 equipment qualification program to ress the aging requirement of IEEE 323. For equipment required to perform a safety-related tion in a high-energy line break environment, the AP1000 equipment qualification program udes an aging simulation as part of its qualification test sequence (Subprogram A of chment B).
equipment not required to perform a safety related function in a high-energy line break ironment, the single design basis event considered is a seismic event. Aging, in this case bprogram B of Attachment B) is not usually included in the test sequence. Aging, where ificant, is addressed by separate qualification of aged components, using conservative testing er applicable seismic design basis event conditions.
rmal effects are one of the primary aging mechanisms addressed by the AP1000 equipment lification program described in Attachment B for equipment containing nonmetallic or nonceramic erials. When thermal aging effects are established as potentially significant to the capability of the ponent or equipment to perform its safety-related function under design basis event conditions, the absence of evidence to the contrary, the component or equipment is thermally aged to ulate an end-of-qualified-life condition before design basis event testing. Equipment required to rate in a high-energy line break environment is also thermally aged to simulate the post-accident ditions consistent with its established functional requirements.
attachment defines the appropriate thermal environments considered for each item of ipment in the AP1000 equipment qualification program and establishes consequent accelerated mal aging parameters for use in the qualification programs.
Arrhenius Model aging mechanism is governed by a single chemical reaction, the rate of which is dependent on perature alone, the Arrhenius equation can be used as the basis for establishing the accelerated g parameters:
E dR
= Ae kT (1) dt re:
E = activation energy (eV) k = Boltzmann's constant (8.617 x 10-5 eV/K)
A = constant factor T = material temperature (K) dR
= reaction rate = aging rate dt 3D-89 Revision 1
R = Be kT t re:
R = change in measured property due to aging t = time for aging effect R to occur B = constant factor e accelerated aging process employed correctly simulates the change in properties due to aging er normal operating or post-accident temperature conditions, then:
R 1 = R 0 (3)
E E kT1 kT0 B t1 e = B to e E T1 T0 Ln t 1 = + Ln t o k T1T0 re:
T1 = accelerated aging material temperature (K) t1 = time at temperature T1 T0 = material temperature under normal operating or post-accident conditions (K) t0 = time at temperature T0 m Equation 3, given an activation energy (E) for the material, the time required at any selected ated temperature can be calculated to simulate the ambient aging effects.
model has been verified to represent the thermal aging characteristics of nonmetallic and non-mic materials and is employed in the AP1000 equipment qualification program to derive elerated thermal aging parameters. The only material dependent parameter input into this model, n establishing the accelerated aging parameters, is the activation energy. This parameter is a ct measure of the chemical reaction rate governing the thermal degradation of the material.
Activation Energy ngle material may have more than one physical property that thermally degrades (for example, ectric strength, flexural strength.) As a consequence, the material exhibits different activation rgies with respect to each property. The activation energy selected is the one that reflects the sical property most significant to the safety-related function performed or the stresses applied to material by the design basis fault(s) considered.
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rgy is not known.
ere an activation energy is not available that reflects the material or component as well as the sical property of interest, a single conservative activation energy is used.
stribution of activation energies (Figure 3D.D-1) was produced by EPRI (Reference 1) based on materials. An independent review of materials used in Westinghouse-supplied equipment is marized in Table 3D.D-1 and plotted in similar form in Figure 3D.D-2. A statistical analysis cates that 95 percent of the activation energies exceed about 0.4 eV from the EPRI data and eV from the Westinghouse data. Based on this information, a value of 0.5 eV is selected for use ughout this program whenever specific activation energies are not available. Employing a low e of activation energy in deriving the accelerated aging parameters causes materials having a activation energy to be overaged with respect to the simulated conditions.
Thermal Aging (Normal/Abnormal Operating Conditions) section establishes the methodology employed and derives a typical set of accelerated aging ameters for equipment in various plant locations.
1 Normal Operation Temperature (T0) etermining the ambient operating temperature (T0) of the component/material/equipment under stigation, the following is considered:
External ambient temperature (Ta)
Temperature rise in cabinet/enclosure (Tr)
Self-heating effects (Tj) re To = Ta + Tr + Tj 1.1 External Ambient Temperature (Ta)
For equipment located in areas supplied by an air-conditioning system, a typical value assumed for (Ta) throughout the qualified life is 68°F (20°C). For air-conditioning systems, two excursions per year to 91°F (33.3°C), each lasting 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, has a negligible additional aging effect.
For equipment located in areas supplied by a ventilation system, a typical value assumed (Ta) throughout the qualified life is 77°F (25°C). Two excursions per year to 122°F (50°C), each lasting 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, has a negligible additional aging effect.
1.2 Temperature Rise in Enclosure (Tr) temperature rise is estimated based on the heat generated (radiative and conductive) by ipment inside or attached to the enclosure. For example, limit switches may be affected by ess heat through the valve. Temperatures measured during test runs may be available. A typical e for temperature rise inside an electronics cabinet is 10°C.
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blished. If the equipment is energized only for short durations, this effect may be determined to egligible. Temperature effects due to the solenoid of an energized valve may be significant (over C). In determining junction temperatures of semiconductor devices, known operating parameters g with the thermal impedance are used. If the power dissipation is not known, a 50 percent rating stress is assumed.
2 Accelerated Aging Temperature (Ti) peratures used for actual accelerated thermal aging tests are determined based on the ipment or component specifications in an attempt to prevent damage from high temperature e and second-order (non-Arrhenius) effects such as the glass transition temperature of plastics.
aximum of 130°C is typically used for electronic component aging, but this is evaluated on a case is. If the device is energized during the accelerated aging process, the self-heating effect as rmined in the preceding section is added to the oven temperature to determine the total aging perature (Ti).
3 Examples of Arrhenius Calculations 3.1 For a Normally Energized Component Aged Energized - The Self-Heating Effect is Added to Both (To) and (Ti):
Conditions: Ta = 25°C, Tr = 10°C Tj = 25°C, eV = 0.5, Aging time = ti Oven temperature = 130°C Qualified life goal = 10 years Therefore To = 25 + 10 + 25 = 60°C = 333K Ti = 130 + 25 + 155°C = 428K t1 = 10e 0.5 (428 333) = 1831 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.966955e-4 months <br /> K (428 x 333) 3.2 For a Normally De-energized Component Aged Energized - the Self-heating Effect is Added Only to Ti:
Conditions: Ta = 25°C, Tr = 10°C Tj = 25°C, eV = 0.5, Aging time = t1 Oven temperature = 130°C Qualified life goal = 10 years Therefore To = 25 + 10 = 35°C = 308K Ti = 130 + 25 + 155°C = 428K 3D-92 Revision 1
Post-Accident Thermal Aging t cases, some safety-related postaccident performance capability is specified by the functional uirements. As a consequence, to qualify equipment to IEEE 323, the effects of post-accident mal aging must be simulated after the high-energy line break test. This section establishes the elerated thermal aging parameters employed in performing this simulation.
1 Post-Accident Operating Temperatures uming continuous operation of containment safeguards systems following an accident, the tainment environment temperature is reduced to the external ambient temperature well within one r for any postulated high-energy line break. However, to allow for possible variations in plant rations following an accident, the limiting design high-energy line break envelope extending to year is indicated by Figure 3D.5-8.
safety-related equipment located inside containment, either the self-heating effects of the rating unit, under post-accident conditions, may be insignificant compared to the heat input from external environment (transmitters, RTDs), or the unit may not be in continuous operation during phase (valve operators). So it may not be necessary to add a specific temperature increment to ount for self-heating of these devices following an accident. The portion of Figure 3D.5-8 that is addressed by DBA testing is then input at T0 into the Arrhenius equation to calculate appropriate elerated aging parameters for post-accident conditions. However, as noted in Section D.4, if the ipment is energized during the aging simulation period, the self-heating effect is added to both To Ti.
2 Accelerated Thermal Aging Parameters for Post-Accident Conditions aging temperature most often used for post-accident simulation is 250°F (121°C). This perature is selected as a maximum for electronic components and is generally used for tests.
ng this value and the conservative activation energy of 0.5 eV, the Arrhenius equation is applied e curve in Figure 3D.5-8 from one day to four months or to one year in small increments of time.
required aging times to simulate these small increments are then summed to yield a total test of 42 days to simulate four months and about 67 days to simulate one year post-accident ration. Including appropriate margin adds four and seven days respectively to the total test time.
activation energy of 0.8 eV is justified, the Arrhenius equation yields 19 days to simulate four ths and 26 days to simulate one year with two days and three days margin to be included in the l test time.
References EPRI NP-1558, Project 890-1, "A Review of Equipment Aging Theory and Technology,"
September 1980.
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Electron Material Volts Melamine-Glass, G5 0.29 Epoxy B-725 0.48 Ester-Glass, GPO-3 0.57 RTV Silicone 0.60 Phenolic-Asbestos, A 0.61 Nylon 33 GF 0.70 Acetal 0.73 Mineral Phenolic 0.74 Silicone Varnish 0.74 Polypropylene 0.81 Polysulfone 0.83 Phenolic-Cotton, C 0.84 Formvar 0.85 Epoxy 0.88 Epoxy Adhesive 0.89 Nylon 0.90 Pressboard 0.91 Kapton 0.93 Silicone 0.94 Phenolic-Asbestos, A 0.94 Cast Epoxy 0.98 Urethane-Nylon 0.99 Phenolic-Glass, G-3 1.01 Polycarbonate 1.01 Phenolic-Paper, X 1.02 Epoxy Wire 1.05 Epoxy-Glass, FR-4 1.05 Varnish Cotton 1.06 PVC 1.08 Ester-Glass, GPO-1 1.09 Cellulose Phenolic 1.10 X-Link Ethylene 1.11 Urethane 1.12 Ester-Glass, GPO-2 1.13 Ester-Nylon 1.14 3D-94 Revision 1
Material Volts Ester-Glass, GPO-1 1.16 32102BK Varnish 1.16 Vulcanized Fiber 1.16 Cellulose Mineral Phenolic 1.17 Mylar 1.18 Cast Epoxy 1.18 32101EV Varnish 1.18 Epoxy 1.18 Silicone 1.18 Phenolic-Paper, XX 1.20 Vulanized Fiber 1.21 Cellulose Phenolic 1.24 Phenolic-Glass, G-3 1.24 Kraft Phenolic 1.25 Neoprene 1.26 Amide-Imide Varnish 1.31 Loctite 75 1.38 Acetyl. Cotton 1.39 Silicone-Asbestos 1.41 Epoxy-Glass, FR-4 1.50 Mylar 1.58 Nomex 1.59 Omega Varnish 1.59 Epoxy-Glass, G-11 1.64 Polythermaleze 1.64 Kraft Paper 1.67 Valox 310SE-0 1.75 Varnished Kraft 1.86 Nomex 1.91 Ester-Glass, GPO-3 2.03 Phenolic-Cotton, C 2.12 Melamine-Glass, G-5 2.18 3D-95 Revision 1
Figure 3D.D-1 equency Distribution of Activation Energies of Various Components/Materials (EPRI Data) 3D-96 Revision 1
Figure 3D.D-2 Frequency Distribution of Activation Energies of Various Components/Materials (Westinghouse Data) 3D-97 Revision 1
3D-98 Revision 1 Purpose following is the methodology used to seismically qualify seismic Category I mechanical and trical equipment for the AP1000 equipment qualification program. Qualification work covered by appendix meets the applicable requirements of IEEE 344-1987 and 382-1996.
Definitions following are definitions of terms unique to or distinct from common industry usage. (See tion E.4.2.)
1 1/2 Safe Shutdown Earthquake 1/2 safe shutdown earth (SSE) is the earthquake level used during seismic testing to seismically safety-related equipment before performing safe shutdown earthquake testing.
2 Seismic Category I Equipment mic Category 1 equipment consists of structures, systems, and components required to stand the effects of the safe shutdown earthquake and remain structurally intact, leak-tight (in e of pressurized systems), and functional to the extent required to perform their safety-related tion.
3 Seismic Category II Equipment mic Category II equipment is that equipment whose continued function is not required, but se failure could reduce the functioning of seismic Category I structures, systems, and ponents to an unacceptable level. Seismic Category II equipment must be capable of maintaining ctural integrity so that a seismic event up to and including an SSE would not cause such a failure.
4 Non-seismic Equipment ipment designated as non-seismic does not require seismic qualification.
5 Active Equipment ipment that must perform a mechanical or electrical operation during or after (or both) the safe tdown earthquake in order to accomplish its safety-related function.
6 Passive Equipment ipment where maintenance of structural or pressure integrity is the only requirement necessary accomplishing its safety-related function.
Qualification Methods section presents a general description of the seismic qualification methods used by AP1000 for seismic qualification of seismic Category I safety-related mechanical and electrical equipment.
ee methods are used: test, analysis, and a combination of the two. The approaches for lification by testing and by analysis are discussed in Section E.5 and Section E.6, respectively.
following discussion covers the conditions under which each approach is used and the general 3D-99 Revision 1
1 Use of Qualification by Testing preferred method for seismic qualification of safety-related Class 1E electrical and tromechanical equipment is seismic testing. The nature of the seismic and vibrational input used ends on where the equipment is used. For equipment mounted so that the seismic environment udes frequency content between 1 and 33 hertz (hard mounted), the seismic test input is tifrequency. For equipment mounted so that seismic ground motion is filtered to contain one dominant structural mode (line mounted), single frequency testing is appropriate. This is the case equipment mounted on piping systems, ductwork, or cable trays.
2 Use of Qualification by Analysis lysis is used for seismic qualification when one of the following conditions is met:
The equipment is too large or the interface support conditions cannot adequately be simulated on the test table.
The only requirement is to maintain structural integrity during a postulated seismic event.
The equipment represents a linear system, or the nonlinearities can conservatively be accounted for in the analysis. This approach is also applicable to the development of the seismic environment, required response spectrum curve, at the mounting location of a component attached to a larger structure when the device is seismically qualified by separate component testing.
The analysis is used to document the seismic similarity of the equipment provided and that previously qualified by testing.
mic qualification of safety-related electrical equipment by analysis alone is not recommended for plex equipment that cannot be modeled to adequately predict its response. Analysis without ing may be acceptable provided structural integrity alone can ensure the design-intended tion.
Requirements 1 Damping ping level of a component or system describes its capability to dissipate vibrational energy ng a seismic event. The damping level used defines the response magnitude of an ideal single ree of freedom linear oscillator when subjected to the specified input as documented by the uired response spectrum (RRS) curve. The significance of the damping value used depends on ther qualification is by testing or analysis.
1.1 Testing ipment qualification by testing involves subjecting the base of the equipment to a representative mic acceleration time history. The response characteristics of the equipment are a function of the rent damping present in the equipment. In this case the damping value used (typically five ent) serves as a convenient means of showing the compliance of the test response spectrum S) with the required response spectrum.
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ally present in the equipment. Unless other documented equipment damping data is available, values specified in Table 3.7.1-1 of Chapter 3 are used.
2 Interface Requirements part of the seismic qualification program, consideration is given to the definition of the clearances ded around the equipment mounted in the plant to permit the equipment to move during a tulated seismic event without causing impact between adjacent pieces of safety-related ipment. This is done as part of seismic testing by measuring the maximum dynamic relative lacement of the top and bottom of the equipment.
en performing qualification by analysis, the relative motion is obtained as part of the analytical lts. These motions are reported in the qualification report and are used to determine the required rance between adjacent pieces of equipment.
ddition, the qualification program takes into account the restraining effect of other interfaces, such ables and conduits attached to the equipment, which may change the dynamic response racteristic of the equipment.
3 Mounting Simulation mounting conditions simulated by analysis or during seismic test are representative of the ipment as-installed mounting conditions used for the AP1000 equipment. When an interfacing cture exists between the safety-related equipment being qualified and the floor or wall at which equipment mounting required response spectrum is specified, its flexibility is simulated as part of qualification program. If this is not done, justification must be provided, demonstrating that the iations in mounting conditions do not affect the applicability of qualification program.
4 1/2 Safe Shutdown Earthquake AP1000 makes use of a small earthquake having the intensity of one-half of the safe shutdown hquake at the safety-related equipment mounting location to simulate the fatigue effects of ller earthquakes that may occur before the postulated safe shutdown earthquake. These small hquakes correspond to the operating basis earthquakes (OBEs) referenced in IEEE 344-1987.
en qualification by testing is used, five of these small earthquakes are used to vibrationally age equipment before the safe shutdown earthquake. When qualification by analysis is used, two shutdown earthquake events are used to simulate the fatigue aging effects. Each event contains eak cycles. These stress cycles are used to verify that the equipment is not subject to failure due w cycle fatigue.
5 Safe Shutdown Earthquake safe shutdown earthquake required response spectrum curve defines the seismic qualification is for each piece of safety-related equipment. The seismic level varies according to the mounting tion of the equipment. When equipment qualification is based on testing, an additional 10 percent acceleration margin is added as specified in IEEE 323-1974.
6 Other Dynamic Loads rodynamic loads are considered as part of the qualification program, where applicable.
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tromechanical equipment. Seismic testing shall be performed and input generated as specified in E 344-1987. The nature of the test input used depends on whether the equipment is hard nted or line mounted. The test program consists of the following elements, as applicable:
ironmental aging, mechanical aging, vibrational aging, and safe shutdown earthquake testing. For e cases where the equipment is also subject to a loss of coolant or a high-energy line break dent, these accidents are simulated on the same qualification specimen after completion of the ing previously discussed. (See Subsections 3D.4.4 and 3D.7.4.)
characteristics of the required seismic and dynamic input motions should be specified by the onse spectrum or time history methods. These characteristics, derived from the structures or ems seismic and dynamic analyses, should be representative of the input motions at the ipment mounting locations.
seismic and dynamic loads, the actual test input motion should be characterized in the same ner as the required input motion, and the conservatism in amplitude and frequency content uld be demonstrated (that is, the test response spectrum should closely resemble and envelop required response spectrum over the critical frequency range).
e seismic and the dynamic load excitation generally have a broad frequency content, multi-uency vibration input motion should be used. However, single frequency input motion, such as beats, is acceptable provided the characteristics of the required input motion indicate that the ion is dominated by one frequency (for example, by structural filtering effects), or that the cipated response of the equipment is adequately represented by one mode, or in the case of ctural integrity assurance, that the input has sufficient intensity and duration to produce ciently high levels of stress for such assurance. Components that have been previously tested to E-344-1971 should be reevaluated or retested to justify the appropriateness of the input motion d, and requalified if necessary.
the seismic and dynamic portion of the loads, the test input motion should be applied to one ical axis and one principal axis (or two orthogonal axes) simultaneously unless it can be onstrated that the equipment response motion in the horizontal direction is not sensitive to the atory motion in the horizontal direction, and vice versa. The time phasing of the inputs in the ical and horizontal directions must be such that a purely rectilinear resultant input is avoided. An eptable alternative is to test with vertical and horizontal inputs in-phase, and then repeat the test inputs 180 degrees out-of-phase. In addition, the test must be repeated with the equipment ted 90 degrees horizontally.
1 Qualification of Hard-Mounted Equipment d-mounted equipment is seismically tested mounted on a test table capable of producing tifrequency, multiaxis inputs. The waveform characteristics of the input are random and scaled in h a way that the test response spectrum equals or exceeds the required response spectrum luding margin). The input signal meets the requirements of Subsection 7.6.3 of IEEE 344-1987.
hermore, the test input simulates the multidirectional nature of the earthquake. The preferred hod for meeting this requirement is to the use a triaxial test table capable of producing three istically independent, orthogonal input motions. In this case the seismic testing consists of 1/2 safe shutdown earthquake tests and one safe shutdown earthquake test in one orientation.
ng a biaxial test table is acceptable if it is justified that the horizontal and vertical test inputs servatively simulate the three-dimensional nature of the seismic event. One acceptable approach 3D-102 Revision 1
es, with the equipment rotated 90 degrees about the vertical axis. In this case, the five 1/2 safe tdown earthquake inputs need to be applied only in the first orientation.
dependent biaxial test table is used, the test is performed in four stages. The first stage involves 1/2 safe shutdown earthquake tests and one safe shutdown earthquake test in the first ntation. The second, third, and fourth orientations are obtained by successively rotating the ipment 90 degrees clockwise from its previous position. One safe shutdown earthquake test is ormed in each of the last three orientations.
h multifrequency test has a minimum of 15 seconds of strong motion input. The strong motion ion is preceded and followed by a period of testing where the test input is ramped up and ramped n, respectively, so that the equipment is not subjected to impact loading. The adequacy of each run is evaluated using the criteria set forth in Subsection 7.6.3.1 of IEEE 344-1987.
2 Qualification of Line-Mounted Equipment
-mounted equipment, because of the dynamic filtering characteristics of its mounting, is ctively subject to single frequency input. This condition is common for valves and sensors ported by piping systems, cable trays, and duct systems. This equipment is qualified consistent the requirements of IEEE 382-1996.
ome cases this equipment may also be used in the hard-mounted condition. In this case tifrequency, multiaxis testing is also required unless justification is provided that the previous le frequency tests demonstrate the capability of the equipment to operate under the d-mounted seismic conditions. Because of the large size of typical valves, it may be necessary to orm separate testing of the operators and valve assembly.
2.1 Seismic Qualification Test Sequence seismic qualification process is broken down into the following steps:
Mount the equipment on a rigid test fixture and perform a resonant search test to demonstrate that the equipment is structurally rigid (fundamental frequency greater than 33 hertz) and does not amplify the seismic motions acting at the equipment mounting interface.
Perform single frequency testing on the line-mounted equipment.
Perform multifrequency, multiaxis testing on the equipment, if appropriate.
If an active valve assembly is to be seismically qualified, additional testing is needed as follows:
- a. Perform a static pull test on the valve.
- b. Perform a static seismic analysis using a verified model of the valve and its extended structure to demonstrate that the valve has adequate structural strength to perform its safety-related function without exceeding the design allowable stresses specified in ASME Code,Section III, Subsection NB, NC, or ND for pressure-retaining parts, as 3D-103 Revision 1
2.2 Line Vibration Aging
-mounted equipment may be subject to operational vibrations resulting from normal plant rations. The potential fatiguing effect of this vibrational aging is simulated as part of the lification program. This requirement is satisfied by subjecting the equipment to a sine sweep from 100 to 5 hertz at an acceleration level of 0.75g or such reduced acceleration at low frequencies mit the double amplitude to 0.025 inch as specified in Section 5.3.a, Part III of IEEE 382-1996.
2.3 Single Frequency Testing single frequency testing acceleration waveform is either sine beat or sine dwell applied at
-third octave frequency intervals as specified in IEEE 382-1996. Each dwell has a time length quate to permit performance of functional testing, with a minimum time of 15 seconds. To account he three-dimensional nature of the seismic event, the test input level is taken as the square root wo times the required input motion (RIM) level specified in IEEE 382. The level includes the ercent test margin. Each test series is performed using single axis input. The test series is ormed successively in each of three orthogonal axes.
2.4 Seismic Aging aging effect of the five 1/2 safe shutdown earthquake earthquakes can be simulated by exposing equipment to two sinusoidal sweeps at one-half of the safe shutdown earthquake required input ion level in each orthogonal axis. Each sweep shall go from 2 to 35 hertz to 2 hertz at a rate not to eed one octave per minute. One sweep is performed with the equipment in its inactive mode, and other with the equipment in its safety-related operational mode.
2.5 Static Deflection Testing of Active Valves seismic testing just discussed is normally performed only on the valve operator and the attached urtenances. If the valve assembly is rigid, the operability of the valve assembly during a tulated seismic event may be demonstrated by performing a static pull test using a peak eleration value equivalent to a triaxial acceleration of 6g. If the valve assembly is determined to lexible, a supplemental analysis of the seismic response of the flexible valve and its supporting ng is performed to determine the actual acceleration level present at the center of gravity of the e assembly.
valve is placed in a suitable test fixture with the operator and appurtenances mounted and nted as in the normal valve assembly installation. The valve is mounted so that the extended cture is freestanding and supported only by the valve nozzles. The valve is positioned so that the zontal and vertical load components simulating the three-dimensional nature of the seismic event duce a worst-case stress condition in the valve extended structure.
ing testing, the valve shall be internally pressurized and nozzle loads applied. Static loads ulating dead weight and seismic loads are applied to the extended structure. The tests are mally performed at ambient temperature. These loads simulate to the extent feasible the load ribution acting on critical parts of the valve assembly. The valve is actuated using the actuator em seismically qualified according to IEEE 382-1996. The valve assembly is cycled from its mal to the desired safety-related position within the time limits defined in the equipment cification. Leakage measurements are made, where required, and compared to the allowable es specified in the valve design specification.
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hquake, the equipment is operated and monitored to demonstrate that the equipment functions perly before, during, and after the seismic event. If the test time is not long enough to complete required functional tests, the length of the strong motion test time is increased to permit pletion of the required functional testing.
ere functional testing is dependent on external electrical supply, the testing is performed using the st-case electrical supply conditions.
4 Resonant Search Testing onant search testing is performed to provide data on the natural frequency and dynamic onse characteristics of the equipment qualified. For hard-mounted equipment being qualified by mic testing, resonant search testing is done to provide additional information but is not required qualification of the equipment. This is an important consideration because frequency testing for d-mounted equipment is normally performed with the equipment mounted on the test table, where amic interaction of the table and the equipment has a significant effect on the measured natural uency.
qualification of line-mounted valve assemblies, it is necessary that the assemblies be rigid. To t this requirement, the assembly mounted to a rigid test fixture so that the frequencies measured indeed representative of the valve assembly. If it is not feasible to provide a rigid fixture, as is y the case when testing such very large valves, as the main steam and feedwater isolation es, additional tests and analyses may be required to determine if the apparent flexibility sured is due to the test fixture or to the characteristic of the valve assembly itself.
e resonant search test data is being generated to verify the accuracy of an analytical modeling nique, the test specimen mounting details must accurately simulate the boundary conditions d in the analytical model.
Qualification by Analysis tion E.3.2 defines the limits on the use of analysis to demonstrate seismic qualification of safety-ted equipment. The following sections describe the analytical methods to be employed for lification of equipment. There are two techniques, static and dynamic, used to qualify equipment.
success of either method depends on the ability of the analytical model to describe the response e system to seismic loads. Alternative methods of analysis are accepted if their conservatism is umented.
analysis is used to demonstrate the structural adequacy of the equipment being qualified. This is e by showing that the calculated stresses do not exceed the design allowable stresses specified SME Code,Section III, Subsection NB, NC, or ND for pressure-retaining equipment and section NF for nonpressure-retaining equipment.
1 Modeling lysis may be performed by hand calculations, finite element, or mathematical models that quately represent the mass and stiffness characteristics of the equipment. The model contains ugh degrees of freedom to adequately represent the dynamic behavior over the frequency range terest. It includes the essential features of the equipment.
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ided that it is insignificant or that the linear model provides conservative results. The adequacy of model or of the modeling techniques is shown by comparing the predicted responses to the onses predicted by benchmark problems or modal testing. Acceptable benchmark problems ude hand calculations, analysis of the same problem using a comparable verified public-domain gram, empirical data, or information from the technical literature.
ddition to documenting the modeling technique, a quality assurance program is in place that nes the requirements for the control, verification, and documentation for the computer programs d for qualification of safety-related equipment. The computer programs used in the qualification ess are verified on the same computer on which the qualification analysis is performed.
2 Qualification by Static Analysis rigid equipment, the seismic forces resulting from one seismic input direction are calculated for h node point by multiplying the nodal mass in that direction by the appropriate zero period eleration (ZPA) floor acceleration. The combined system response of the equipment to the ultaneous loads acting in all three directions is calculated by combining the three components, g the square root sum of the squares (SRSS) method. The square root sum of the squares hod is used to account for the statistical independence of the individual orthogonal seismic ponents.
3 Qualification by Dynamic Analysis e lowest natural frequency of the equipment lies below the cutoff frequency, the response of the ipment to the seismic event in each orthogonal direction will be dynamically amplified and the ipment is said to be flexible. The analysis is performed in compliance with the guidelines set forth e SSAR and in Regulatory Guides 1.92, 1.100, and 1.122.
preferred method of analysis is the response spectrum method. In this method the responses in h equipment mode are calculated separately and combined by the square root sum of the ares method, provided the modes are not closely spaced. (Consecutive modes are said to be ely spaced if their frequencies differ from that of the first mode in the group by less than ercent.) The responses for each mode in a group are combined absolutely. The group response en combined with the remaining modal responses using the square root sum of the squares hod. The responses for each of the three orthogonal seismic components can then be combined iscussed in Section E.6.2. The applicable damping levels are noted in Table 3.7.1-1 of Chapter 3.
3.1 Response Spectrum Analysis es up to and including the cutoff frequency are included in this summation. In some cases, the cture is basically rigid, with some of the flexible mode representing local effects. This situation is luated by reviewing the modal masses applicable to a given seismic input direction. If the sum of effective modal masses used in the response spectrum analysis is greater than 0.9 times the total ipment mass, the model is assumed to adequately represent the total equipment mass. If this rion is not satisfied, it means that a significant part of the equipment seismic response is due to static seismic response of the higher equipment modes (above the cutoff frequency). If this ation occurs, the analyst determines the component of the response due to the higher modes and bines it with the flexible response component by square root sum of the squares. (This uirement is discussed in the SSAR, Subsection 3.7.2.)
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- d. In this method the frequencies of the equipment are not determined, but a static analysis is ormed, assuming that a peak acceleration equal to 1.5 times the peak spectral acceleration given e applicable required response spectrum acts on the structure as described in Section E.6.2.
static coefficient of 1.5 takes into account the combined effects of multifrequency excitation and timode response for equipment and structures that can be represented by a simple model. A er static coefficient may be used when it can be demonstrated that it will yield conservative lts.
3.3 Time History Analysis time-history method of analysis is the preferred method of analysis when the equipment exhibits ificant nonlinear behavior or when it is necessary to generate response spectra for specific ponent mounting locations in the equipment. The acceptable methods that are used to develop seismic time histories are discussed in Regulatory Guide 1.122, ASME Code,Section III, endix N, and in Section 6.2 of IEEE 344-1987. Other analytical methods may be used to erate in-equipment response spectra provided that they are verified to produce accurate and/or servative results.
Qualification by Test Experience method of qualification is not used.
Performance Criteria 1 Equipment Qualification by Test performance criterion for qualification of equipment is that the equipment successfully perform its ty-related function during and after the postulated seismic event. Acceptance requires, as a imum, that:
No spurious or unwanted outputs occur in the circuits that could impair the safety-related functional operability of the equipment; No gross structural damage of the equipment occur during the seismic event that could lead to the equipment or any part thereof becoming a missile. Local inelastic deformation of the equipment is permitted; and, Satisfactory completion of specified baseline tests are demonstrated before, during, and after the seismic test sequence.
2 Equipment Qualification by Analysis 2.1 Structural Integrity analysis verifies that the equipment, when subjected to the worst case combination of operating seismic loads, maintains its structural integrity. In addition the analysis shows that the equipment ot subject to low cycle fatigue failure when subject to postulated seismic loading. Finally the lysis verifies that seismically induced equipment motion does not lead to impacting with other rby equipment.
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ctural integrity or limitation of deformation guarantees operability. As an example the analysis of ve equipment verifies that the combination of operating and postulated seismic loads do not duce stress levels or deformations that exceed established functional limits. The rationale for use ese limits is justified.
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didate leak-before-break piping is identified in Figures 3E-1 through 3E-5 along with other piping which high-energy pipe failures are postulated. These figures also identify piping in the break usion zones inside and outside containment. These figures do not include piping of 1 inch size smaller. Instrumentation and instrumentation lines are not included.
selection of the failure type is based on whether the system is high or moderate energy during mal operating conditions of the system. High-energy piping includes those systems or portions of ems in which the maximum normal operating temperature exceeds 200°F or the maximum mal operating pressure exceeds 275 psig. Piping systems or portions of systems pressurized ve atmospheric pressure during normal plant conditions and not identified as high energy are sidered moderate energy. Piping systems that exceed 200°F or 275 psig for 2 percent or less of time during which the system is in operation or that experience high-energy pressures or peratures for less than 1 percent of the plant operation time are considered moderate energy. In ng whose nominal diameter is greater than 1 inch but less than 4 inches, only circumferential aks are postulated at each selected location. No breaks are postulated for piping whose nominal meter is 1 inch or less.
three-letter code included in the line numbering identifies the pipe specification. The letters ne the pressure class, material specification, and AP1000 equipment classification, respectively.
symbols used in Figures 3E-1 through 3E-5 are the same as the P&ID figures. See Figure 1.7-2 dditional information on the drawing legend and for the key for the pipe specification. Section 3.2 udes additional information on the AP1000 equipment classification.
3E-1 Revision 1
WLS 1&2 - UFSAR Figure 3E-1 (Sheet 1 of 2)
High Energy Piping - Steam Generator System 3E-2 Revision 1
WLS 1&2 - UFSAR Figure 3E-1 (Sheet 2 of 2)
High Energy Piping - Steam Generator System 3E-3 Revision 1
WLS 1&2 - UFSAR Figure 3E-2 High Energy Piping - Normal Residual Heat Removal System 3E-4 Revision 1
WLS 1&2 - UFSAR Figure 3E-3 (Sheet 1 of 2)
High Energy Piping - Reactor Coolant System 3E-5 Revision 1
WLS 1&2 - UFSAR Figure 3E-3 (Sheet 2 of 2)
High Energy Piping - Reactor Coolant System 3E-6 Revision 1
WLS 1&2 - UFSAR Figure 3E-4 (Sheet 1 of 2)
High Energy Piping - Passive Core Cooling System 3E-7 Revision 1
WLS 1&2 - UFSAR Figure 3E-4 (Sheet 2 of 2)
High Energy Piping - Passive Core Cooling System 3E-8 Revision 1
WLS 1&2 - UFSAR Figure 3E-5 (Sheet 1 of 2)
High Energy Piping - Chemical and Volume Control System 3E-9 Revision 1
WLS 1&2 - UFSAR Figure 3E-5 (Sheet 2 of 2)
High Energy Piping - Chemical and Volume Control System 3E-10 Revision 1
mic Category II cable trays and their supports are also designed utilizing the design criteria of appendix.
Codes and Standards design of cable trays and their supports conform to the following codes and standards:
American Iron and Steel Institute (AISI), Specification for the Design of Cold Formed Steel Structural Members, 1996 Edition and Supplement No. 1, July 30, 1999 American Institute of Steel Construction (AISC), Specification for the Design, Fabrication and Erection of Steel Safety Related Structures for Nuclear Facilities, AISC-N690-1994 Institute of Electrical and Electronic Engineers (IEEE), Standard 344-1987, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations National Electrical Manufacturers Association (NEMA), Standard Publication No. VE 1-1998, Metallic Cable Tray Systems Loads and Load Combinations
.1 Loads
.1.1 Dead Load (D) d load includes the weight of the cable trays, their supports and the cables inside the trays and permanently attached items. Temporary items used during construction or maintenance are oved prior to operation.
so includes the weight of Cable tray covers and Other components and fittings
.1.2 Construction Live Load (L) load consists of a load of 250 pounds to be applied only during construction on the tray at a cal location to maximize flexural and shear stresses. This load is not combined with seismic s.
.1.3 Safe Shutdown Earthquake (Es) mic response of the cable trays and their supports are produced due to seismic excitation of the ports.
.1.4 Thermal Load se loads are usually not considered and trays are provided with expansion joints in accordance NEMA.
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D + L D + Es Analysis and Design le trays and their supports are designed to maintain structural integrity. The stresses are ntained within the allowable limits as specified in Subsection 3F.3.3. Section properties and ghts of the trays are obtained from manufacturer's data.
.1 Damping maximum damping ratio is 10 percent unless the configuration is demonstrated to be similar to of the tests described in (Reference 19) of Subsection 3.7.6.
tated in Subsection 3.7.1.3, the damping ratio used for the AP1000 cable tray systems may be ed on test results presented in Reference 19 (Subsection 3.7.6). The cable tray test program ducted by ANCO Engineers Inc. included more than 2000 dynamic tests of representative cable system design and construction. The test configurations included items such as various tray s on rigid supports, various tray hanger systems, effects of tray types, effects of strut connections effects of bracing spacing, unbraced and braced tray systems. Cable ties were also used during test program. Based on observations during the tests, the high damping values within the cable system are provided mainly by the movement, sliding or bouncing of the cables within the tray.
tests show that, for unloaded trays, the damping ratio closely approximates the 7 percent used bolted structures, and a minimum damping value of 20 percent is maintained with cable ties at cing greater than or equal to four feet. The tests show that for loaded trays, the damping ratio eases with increased cable loading, reaching a value of 30 percent at cable fill ratio of 50 percent 00 percent. The major factors which affect the damping ratio of the cable tray systems are the t acceleration level, cable fill ratio, and the ability of the cables to move within the trays during a shutdown earthquake.
AP1000 cable tray system design requires no sprayed-on material for fire protection. Cable ties provided at spacing greater than 4 feet, thereby permitting cable movement within the trays. The ping ratio used for the cable tray system is dependent on the level of seismic input and the unt of cable fill within the trays. As shown in Figure 3.7.1-13, the 20 percent constant damping may be used for trays loaded to more than 50 percent and subjected to input floor acceleration ater than 0.35g. For cable trays loaded to less than 50 percent and lower than 0.35g input floor eleration, linearly interpolated lower damping values may be used.
.2 Seismic Analysis methodology for seismic analysis is provided in Subsection 3.7.3. Seismic loads are determined ither using the equivalent static load method of analysis or by performing dynamic analysis.
sses are determined for the seismic excitation in two horizontal and one vertical direction. The sses in the three directions are combined using the square root of the sum of the squares (SRSS) hod or the 100-40-40 method as described in Subsection 3.7.3.6.
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cification. The basic stress allowables for cable tray supports utilizing light gage cold rolled nnel type sections are based on the manufacturer's published catalog values. The basic stress wables for cable tray supports utilizing rolled structural shapes are in accordance with SI/AISC N-690 and the supplemental requirements described in Subsection 3.8.4.5.2.
allowable stresses for the load combinations are as follows:
+L Basic Allowable Es 1.6 times basic allowable for tension and 1.4 times basic allowable for compression
.4 Connections nections are designed in accordance with the applicable codes and standards listed in tion 3F.1. For connections used with light gage cold rolled channel type sections, design is based he manufacturer's published catalog values. Supports are attached to the building structure by ed or welded connections. Fastening of the supports to concrete structures meets the plemental requirements given in Subsection 3.8.4.5.1.
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appendix summarizes the seismic analyses of the nuclear island building structures performed upport the AP1000 design certification extension from just hard rock sites, to sites ranging from soils to hard rock. The seismic Category I building structures consist of the containment building steel containment vessel [SCV] and the containment internal structures [CIS]), the shield ding, and the auxiliary building. These structures are founded on a common basemat and are ectively known as the nuclear island or nuclear island structures. Key dimensions of the seismic egory I building structures, such as thickness of the basemat, floor slabs, roofs and walls, are wn in Figures 3.7.1-14 and 3.7.2-12.
lyses were performed in accordance with the criteria and methods described in Section 3.7.
tion 3G.2 describes the development of the finite element models. Section 3G.3 describes the soil cture interaction analyses of a range of site parameters and the selection of the parameters used e design analyses. Section 3G.4 describes the fixed base and soil structure interaction dynamic lyses and provides typical results from these dynamic analyses. References 3 and 6 provide a mary of dynamic and seismic analysis results (i.e., modal model properties, accelerations, lacements, response spectra) and the nuclear island liftoff analyses. The seismic analyses of the lear island are summarized in a seismic analysis summary report. Deviations from the design due s-procured or as-built conditions are acceptable based on an evaluation consistent with the hods and procedures of Sections 3.7 and 3.8 provided the following acceptance criteria are met:
The structural design meets the acceptance criteria specified in Section 3.8.
The seismic floor response spectra (FRS) meet the acceptance criteria specified in Subsection 3.7.5.4.
ending on the extent of the deviations, the evaluation may range from documentation of an ineering judgment to performance of a revised analysis and design. The results of the evaluation be documented in an as-built summary report by the Combined License applicant.
le 3G.1-1 and Figure 3G.1-1 summarize the types of models and analysis methods that are used e seismic analyses of the nuclear island, as well as the type of results that are obtained and re they are used in the design. Table 3G.1-2 summarizes the dynamic analyses performed and methods used for combination of modal responses and directional input.
2 Nuclear Island Finite Element Models AP1000 nuclear island consists of three distinct seismic Category I structures founded on a mon basemat. The three building structures that make up the nuclear island are the coupled iliary and shield building (ASB), the SCV, and the CIS. The shield building and the auxiliary ding are monolithically constructed with reinforced concrete and, therefore, considered one cture. The nuclear island is embedded approximately 40 feet with the bottom of basemat at ation 60-6 and plant grade located at elevation 100-0. The CSV is described in section 3.8.2, the CIS in Subsection 3.8.3, the ASB in Subsection 3.8.4, and the nuclear island emat in Subsection 3.8.5.
mic systems are defined, according to SRP 3.7.2 (Reference 1),Section II.3.a, as the seismic egory I structures that are considered in conjunction with their foundation and supporting media to a soil-structure interaction model. Fixed base seismic analyses are performed for the nuclear nd at a rock site. Soil-structure interaction analyses are performed for soil sites. The analyses erate a set of in-structure responses (design member forces, nodal accelerations, nodal 3G-1 Revision 1
e value to reduce stiffness to simulate cracking.
ismic response spectrum analysis is performed to develop the seismic design loads for the ign of the auxiliary building, shield building, and containment internal structure, and the loads erated include the amplified load due to flexibility and the distribution of this load to the ounding structures. Equivalent static analyses are used to design the shield building roof and al roof beams, tension ring, air inlet structure, and PCS tank.
2.1 Individual Building and Equipment Models 2.1.1 Coupled Auxiliary and Shield Building finite element shell dynamic model of the coupled ASB is a finite element model using primarily ll elements. The portion of the model up to the elevation of the auxiliary building roof is developed g the solid model features of ANSYS, which allow definition of the geometry and structural perties. The nominal element size in the auxiliary building model is about 9 feet so that each wall two elements for the wall height of about 18 feet between floors. This mesh size, which is the e as that of the solid model, has sufficient refinement for global seismic behavior. It is combined a finite element model of the shield building roof and cylinder above the elevation of the auxiliary ding roof. This model is shown in Figure 3G.2-1. This finite element shell dynamic model is part of NI10 model.
e the water in the passive containment cooling system tank responds at a very low frequency shing) and does not affect building response, the passive containment cooling system tank water s is reduced to exclude the low frequency water sloshing mass. The wall thickness of the bottom ion of the shield building (elevation 63.5 to 81.5) is modeled as one half (1.5) since the CIS el is connected to this portion and extends out to the mid-radius of the shield building cylindrical
. Local portions of the ASB floors and walls are modeled with sufficient detail to give the response e flexible areas.
2.1.2 Containment Internal Structures finite element shell model of the containment internal structures is a finite element model using arily shell elements for the walls and floors and solid elements for the mass concrete. It is eloped using the solid model features of ANSYS, which allow definition of the geometry and ctural properties. This model is used in both static and dynamic analyses. It models the inner and r mass concrete basemats embedding the lower portion of the containment vessel, and the crete structures above the mass concrete inside the containment vessel. The walls and basemat de containment for this model are shown in Figure 3G.2-2. The basemat (dish) outside the tainment vessel is shown in Figure 3G.2-3. This finite element shell dynamic model is part of the 0 model. Static analyses are also performed on the model to obtain member forces in the walls.
model is also used in the 3D finite element basemat model (see Subsection 3.8.5.4.1).
2.1.3 Containment Vessel SCV is a freestanding, cylindrical, steel shell structure with ellipsoidal upper and lower steel es. The finite element model of the containment vessel is an axisymmetric model fixed at ation 100. Static analyses are performed with this model to obtain shell stresses as described in section 3.8.2.4.1.1. The model is also used to develop modal properties (frequencies and mode pes). The three-dimensional, lumped-mass stick model of the SCV is developed based on the 3G-2 Revision 1
Members representing the cylindrical portion are based on the properties of the actual circular cross section of the containment vessel.
Members representing the bottom head are based on equivalent stiffnesses calculated from the shell of revolution analyses for static 1.0g in vertical and horizontal directions.
Shear, bending and torsional properties for members representing the top head are based on the average of the properties at the successive nodes, using the actual circular cross section.
These are the properties that affect the horizontal modes. Axial properties, which affect the vertical modes, are based on equivalent stiffnesses calculated from the shell of revolution analyses for static 1.0g in the vertical direction.
equivalent static acceleration analyses of the containment vessel use a finite element shell el with a refined mesh in the area adjacent to the large penetrations. Comparison of this with a history analysis for the regions immediately surrounding the large penetrations verifies that the s from equivalent static analysis are conservative to time history using a representative study.
stick model is combined with the polar crane stick model as shown in Figure 3G.2-4. Modal perties of the containment vessel with and without the polar crane are shown in Table 3G.2-1. It is nected to nodes on the dish model. NI10 node numbers are shown in red and NI20 node bers are shown in black.
method used to construct a stick model from the axisymmetric shell model of the containment sel is verified by comparison of the natural frequencies determined from the stick model and the ll of revolution model as shown in Table 3G.2-2. The shell of revolution vertical model 0 harmonic) has a series of local shell modes of the top head above elevation 265 between nd 30 hertz. These modes are predominantly in a direction normal to the shell surface and not be represented by a stick model. These local modes have small contribution to the total onse to a vertical earthquake as they are at a high frequency where seismic excitation is small.
only seismic Category I components attached to this portion of the top head are the water ribution weirs of the passive containment cooling system. These weirs are designed such that r fundamental frequencies are outside the 23 to 30 hertz range of the local shell modes.
evaluation was made of the connection of the bottom of the steel containment vessel stick model e CIS finite element model. Comparisons were made between the unconstrained fully metric, radially constrained fully symmetric, and original asymmetric connectivity models. The onse spectra at the elevation of the polar crane girder for the first two models are almost tical, and the third model had only minor differences. Based on this comparison, the onstrained fully symmetric connectivity model is used.
2.1.4 Polar Crane polar crane is supported on a ring girder, which is an integral part of the SCV at elevation 228-0, hown in Figure 3.8.2-1. It is modeled as a multi-degree of freedom system attached to the steel tainment shell at elevation 224 (midpoint of ring girder) as shown in Figure 3G.2-4. The polar e is modeled using a simplified and detailed model. The simplified model has five masses at the
-height of the bridge at elevation 233-6 and one mass for the trolley, as shown in re 3G.2-5A. The polar crane model includes the flexibility of the crane bridge girders and truck embly, and the containment shells local flexibility. When fixed at the center of containment, the el shows fundamental frequencies of 3.3 hertz transverse to the bridge, 7.0 hertz vertically, and hertz along the bridge. The Detailed Model of the polar crane consists of 28 nodes is defined 3G-3 Revision 1
es 1 to 4 represent the Trucks with elevation at top of rails (TOR). There are four nodes that are cident with nodes 1 to 4 and used to add the local SCV stiffnesses (nodes 465 to 468, not shown igure).
Nodes 9 to 12 represent the trolley. The trolley is connected to the centerline of the polar crane girders at nodes 9 and 10.
Nodes 13 to 26 are located on the polar crane girders. The end nodes (13, 19, 20 and 26) are used to connect the cross beams to the girders; these nodes are also attached to the trucks (nodes 1 to 4) by rigid links.
Node 470 is at the center of containment at the top of rail elevation. Nodes 465 to 468 are attached to node 470 using rigid links.
Node 29, not shown in Figure, is located on the SCV. It is attached to 470 by a rigid link.
2.1.5 Major Equipment and Structures Using Stick Models major equipment supported by the CIS is represented by stick models connected to the CIS.
se stick models are the reactor coolant loop model shown in Figure 3G.2-6, the pressurizer model wn in Figure 3G.2-7, and the core makeup tank model shown in Figure 3G.2-8. The core makeup model is used only in the nuclear island fine (NI10) model; the core makeup tank is represented mass in the nuclear island coarse model (NI20).
2.2 Nuclear Island Dynamic Models te element shell models (3D) of the nuclear island concrete structures are used for the time ory seismic analyses. Stick models are coupled to the shell models of the concrete structures for containment vessel, polar crane, the reactor coolant loop and pressurizer. Two models are used.
fine (NI10) model is used to define the seismic response for the hard rock site. The coarse
- 0) model is used for the soil structure interaction (SSI) analyses. It is similar to the NI10 model the exception that the mesh size for the ASB and CIS is approximately 20 feet instead of 10 feet.
model is set up in both ANSYS and SASSI. The NI05 model is used to develop amplified mic response for the envelope of soil profiles presented in Subsection 3.7.1.4 for flexible regions captured by the coarser NI20 model. The NI05 model is also used in response spectrum analysis e nuclear island to develop design seismic member forces and moments. The NI10, NI20, and 5 models are described in the subsections below.
2.2.1 NI10 Model large solid-shell finite element model of the AP1000 nuclear island shown in Figure 3G.2-9 bines the ASB solid-shell model described in Subsection 3G.2.1.1, and the CIS solid-shell model cribed in Subsection 3G.2.1.2. The containment vessel and major equipment that are supported he CIS are represented by stick models and are connected to the CIS. These stick models are SCV and the polar crane models, the reactor coolant loop model, core makeup tank models, and pressurizer model. The stick models are described in Subsections 3G.2.1.3 and 3G.2.1.4. The and attached sticks are shown in Figure 3G.2-10. This AP1000 nuclear island model is referred s the NI10 or fine model. The ASB portion of this model has a mesh size of approximately 10 feet.
SCV is connected to the CIS model using constraint equations. The SCV at the bottom of the k at elevation 100 (node 130401) is connected to CIS nodes at the same elevation. Figure 3G.2-4 3G-4 Revision 1
tions of node 130401. The CIS tangential displacement is tied rigidly (constrained) to the zontal displacement and RZ rotation of node 130401.
2.2.2 NI20 Model NI20 coarse model has fewer nodes and elements than the NI10 model. It captures the essential ures of the nuclear island configuration. The nominal shell and solid element dimension is about eet. It is used in the soil-structure interaction analyses of the nuclear island performed using the gram SASSI. The stick models are the same as used for the NI10 model except that the core eup tank is not included. This model is shown in Figures 3G.2-11 and 3G.2-12. Results of fixed e analyses of the NI20 model were compared to those of the NI10 model to confirm the adequacy e NI20 model for use in the soil-structure-interaction analyses.
2.2.3 Nuclear Island Stick Model nuclear island lumped-mass stick model consists of the stick models of the individual buildings rconnected by rigid links. Each individual stick model is developed to match the modal properties e finite element models described in Subsections 3G.2.1.1 and 3G.2.1.2 above. Modal analyses seismic time history analyses were performed using this model for the hard rock design ification.
nuclear island lumped-mass stick model has been replaced in the design analyses described in appendix by the NI10 and NI20 finite element shell dynamic models of the nuclear island cribed in Subsections 3G.2.2.1 and 3G.2.2.2 above. A 2D stick model is used in the soil sensitivity lyses described in Section 3G.3.
2.2.4 NI05 Model NI05 solid-shell finite element model of the AP1000 nuclear island is shown in Figures 3G.2-13 G.2-15. The NI05 model is used for response spectrum analysis of the nuclear island auxiliary shield building structures. The NI05 model is also used for the mode superposition time history lysis of the nuclear island for the amplified response at flexible floors. The NI05 model is used for static analysis of the nuclear island for the basemat design. The NI05 model is a refined version e NI10 model where the auxiliary and shield building mesh size is reduced from approximately eet by 10 feet tetrahedral mesh to approximately 5 feet by 5 feet. The major equipment stick els supported by the CIS are the same as used for the NI10 model. The steel containment vessel k model and connections are also the same as the NI10 model. The only difference between the 5 CIS and NI10 CIS is the basemat (bowl) and dish region as shown in Figure 3G.2-15. The el is validated by a comparison of the mass participation by frequency of the fundamental modes ose of the NI10 model.
2.2.5 Seismic Stability Model sliding stability of the nuclear island basemat is evaluated using a non-linear 2D East-West (EW) k model of the nuclear island structures using the ANSYS program. Three concentric sticks esent ASB, CIS, and SCV, respectively. The reactor coolant loop is included as mass only. The emat is modeled as a rigid beam, which is free in translation along the EW and vertical directions.
nuclear island combined sticks are attached to the rigid basemat at the nuclear island mass ter.
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ion forces between basemat bottom and foundation media as well as foundation media nesses. The friction coefficient between the basemat bottom and the soil media is set at 0.55.
re 3G.2-19 shows the schematic of this non-linear 2D EW nuclear island stick model. The tact elements are free to uplift when the upward force (normal force) is larger than the associated d load component. When the tangential force is larger than the friction force, sliding occurs.
2.3 Static Models mber forces in the ASB are obtained from analyses of a model that is more refined than the finite ment model described in Subsection 3G.2.1.1. This model is developed by meshing one area of solid model with four finite elements. The nominal element size in this auxiliary building model is ut 4.5 feet so that each wall has four elements for the wall height of about 18 feet between floors.
finite element shell model is referred to as the NI05 model. This refinement is used to calculate design member forces and moments using response spectra analysis of the nuclear island els with seismic input enveloping all soil conditions. The finite element shell model of the tainment internal structures described in Subsection 3G.2.1.2, which includes the basemat within shield building and the containment vessel stick model, is also included.
te element solid/shell models were used for the equivalent static seismic analysis. For the iled design of the shield building roof, a finite element model of one quadrant of the roof is used escribed in Subsection 3G.2.3.1. For the detailed design of the steel containment vessel, a shell h finite element model with a much finer mesh in the areas surrounding the major penetrations is d as described in Subsection 3G.2.3.2. For the static analysis of the containment vessel, an ymmetric model is used as described in Subsection 3G.2.3.3. The nuclear island basemat is luated using the NI05 finite element model described in Subsection 3G.2.2.4.
2.3.1 Quadrant Model of Shield Building Roof one quadrant model of the shield building roof is shown in Figure 3G.2-16. The model is structed with solid and shell elements and contains structures from the exposed shield wall ugh the top of the shield building roof. The quadrant model is used for the equivalent static lysis of the shield building roof. The results from the more detailed analysis are used in the design e shield building roof and radial roof beams, tension ring, air inlet structure, and PCS tank.
2.3.2 Containment Vessel 3D Finite Element Model 3D finite element model of the steel containment vessel is shown in Figure 3G.2-17. The finite ment model for the steel containment vessel is used for the stress analysis of the large etrations (personnel locks and equipment hatches) of the containment vessel.
2.3.3 Containment Vessel Axisymmetric Model axisymmetric finite element model of the steel containment vessel is shown in Figure 3G.2-18.
axisymmetric model is a two-dimensional model with added mass for the stiffeners, crane girder, ipment hatches, and air locks.
3 2D SASSI Analyses section describes the soil structure interaction analyses performed using 2D models in SASSI to ct the design soil cases for the AP1000. The AP1000 footprint, or interface to the soil medium, is tical to the AP600. The AP1000 containment and shield building are 25 6 and 20 6 3G-6 Revision 1
lyses were performed using 2D stick models of the AP1000 for horizontal seismic input with and out adjacent structures. The soil profiles included a hard rock site, a firm rock site, a soft rock a soft-to-medium soil site, an upper bound soft-to-medium site, and a soft soil site. Analyses e also performed without adjacent structures for a hard rock site, a firm rock site, a soft rock site, ft-to-medium soil site, an upper bound soft-to-medium site, and a soft soil site. The soil damping degradation curves are described in Subsection 3.7.1.4. The soil profiles selected for the 000 use the same parameters on depth to bedrock, depth to water table, and variation of shear e velocity with depth as those used in the AP600 design analyses. The Poissons ratio is 0.25 for sites (hard and firm rock) and 0.35 for soil sites (soft-to-medium soil, and upper bound soft-to-ium soil). For all the soil profiles defined, the base rock has been taken to be at 120 feet below de level. The soil profiles are shown in Figure 3G.3-1. The shear wave velocity profiles and related erning parameters are as follows:
For the hard rock site, an upper bound case for rock sites using a shear wave velocity of 8000 feet per second.
For the firm rock site, a shear wave velocity of 3500 feet per second to a depth of 120 feet, and base rock at the depth of 120 feet.
For the soft rock site, a shear wave velocity of 2400 feet per second at the ground surface, increasing linearly to 3200 feet per second at a depth of 240 feet, and base rock at the depth of 120 feet.
For the upper bound soft-to-medium soil site, a shear wave velocity of 1414 feet per second at ground surface, increasing parabolically to 3394 feet per second at 240 feet, base rock at the depth of 120 feet, and ground water at grade level. The initial soil shear modulus profile is twice that of the soft-to-medium soil site.
For the soft-to-medium soil site, a shear wave velocity of 1000 feet per second at ground surface, increasing parabolically to 2400 feet per second at 240 feet, base rock at the depth of 120 feet, and ground water is assumed at grade level.
For the soft soil site, a shear wave velocity of 1000 feet per second at ground surface, increasing linearly to 1200 feet per second at 240 feet, base rock at the depth of 120 feet, and ground water is assumed at grade level.
analyses with and without adjacent structures demonstrated that the effect of adjacent buildings he nuclear island response is small. Based on this, the 3D SASSI analyses of the AP1000 lear island can be performed without adjacent buildings similar to those performed for the AP600.
maximum acceleration values obtained from the AP1000 analyses without adjacent structures given in Table 3G.3-1. The soil cases giving the maximum response are shown in bold. Floor onse spectra associated with nodes 41, 120, 310, 411, and 535 for the six AP1000 soil cases are wn in Figures 3G.3-2 to 3G.3-11.
ed on review of the above results, five soil conditions were selected for 3D SASSI analyses in ition to the hard rock condition evaluated in the existing AP1000 Design Certification. Thus, the wing five soil and rock cases identified in Subsection 3.7.1.4 are considered: hard rock, firm rock, rock, soft-to-medium soil, upper bound soft-to-medium, and soft soil.
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NI10 model described in Subsection 3G.2.2.1 was analyzed by time history modal superposition.
erform the time history analysis of this large model, the ANSYS superelement (substructuring) niques were applied. Substructuring is a procedure that condenses a group of finite elements one element represented as a matrix. The reasons for substructuring are to reduce computer of subsequent evaluations. Two sets of analyses were performed. To obtain the time history onse of the ASB, the ASB finite element model was merged with the superelement of the CIS its major components. To obtain the time history response of the CIS, the CIS finite element el was merged with the superelement of the ASB.
ection time history responses were obtained at selected representative locations. These tions included major wall and floor intersections and nodes at the cardinal orientations at key ations of the shield building. Nodes were also selected at mid-span on flexible walls and floors.
ical locations are shown for the ASB at elevation 135 on Figures 3G.4-1 and 3G.4-2.
re 3G.4-1 shows the rigid locations, and Figure 3G.4-2 shows the flexible locations.
SYS is used to calculate the maximum relative deflection to the nuclear island for the envelope e that considers all of the soil and hard rock site cases. Synthesized displacement time histories developed using the envelope seismic response spectra from the six site conditions (hard rock, rock, soft rock, upper-bound-soft-to-medium, soft-to-medium, and soft soil). Seismic response ctra at nine locations are used (four edge locations, one center location, and four corner tions). It is not necessary to adjust for drift since relative deflections to the basemat are ulated and the drift would be subtracted from the results.
4.2 3D SASSI Analyses computer program SASSI 2000 is used to perform Soil-Structure Interaction analysis with the 0 Coarse Finite Element Model. The SASSI Soil-Structure Interaction analyses are performed for five soil conditions established from the AP1000 2D SASSI analyses. These soil conditions are rock, soft rock, soft-to-medium soil, upper bound soft-to-medium, and soft soil. The model udes a surrounding layer of excavated soil and the existing soil media as shown in Figures 3G.4-3 3G.4-4. Acceleration time histories and floor response spectra are obtained. Adjacent structures e a negligible effect on the nuclear island structures and, thus, are not considered in the 3D SI analyses.
stinghouse has adopted the approach that calculates displacements internally within the ACS SI program based on an analytical complex frequency domain approach that uses inverse Fast-rier Transforms (FFT) to compute relative displacement histories instead of double numerical gration in the time domain that computes absolute displacement time histories from absolute eleration time histories.
relative displacement time history is calculated using ACS SASSI RELDISP module. The plex acceleration transfer functions (TF) are computed for reference and all selected output es. The relative acceleration transfer function is calculated by subtracting the reference node TF the output node TF. The relative displacement transfer function is obtained by dividing the ular frequency square (²) for each frequency data point. The relative displacement time history btained by taking the inverse FFT.
ative displacements are calculated between adjacent buildings and the nuclear island using soft ngs between the buildings. The spring stiffness is very small so that it does not affect the dynamic 3G-8 Revision 1
ese analyses, the three components of ground motions (N-S, E-W, and vertical direction) are t separately. Each design acceleration time history (N-S, E-W, and vertical) is applied separately, the time history responses are calculated at the required nodes. The resulting co-linear time ory responses at a node due to the three earthquake components are then combined braically.
4.3 Seismic Analysis 4.3.1 Response Spectrum Analysis response spectrum methodology used in the AP1000 design employs the Complete Quadratic bination (CQC, Section 1.1 of Reference 5) grouping method for closely spaced modes with the Kiureghian Correlation Coefficient (Section 1.1.3 of Reference 5) used for correlation between es. The Lindley-Yow (Section 1.3.2, Reference 5) spectra analysis methodology is employed for es with both periodic and rigid response components. The modal analysis performed to develop posite modal participation is used to develop input for the response spectrum analysis. Modes ging from 0 to 33 Hz or higher are considered. For modes above the cutoff frequency, the Lindley-is used. The Static ZPA Method (Section 1.4.2, Reference 5) is employed for the residual rigid onse component for each mode as outlined in NRC Regulatory Guide 1.92 (Reference 5). The plete solution is developed via Combination Method B (Section 1.5.2, Reference 5). The bined effects, considering three spatial components of an earthquake (N-S, E-W, and Vertical),
combined by square root sum of the squares method (Section 2.1, Reference 5).
ubsection 3.7.2.6, Three Components of Earthquake Motion, the combination of three ponents of earthquake motion is discussed.
4.3.2 Absolute Accelerations seismic analyses results, which include the new shield building configuration described in tion 3.8, are given in Reference 3.
4.3.3 Seismic Response Spectra AP1000 plant floor response spectra for the six key locations are provided in Figure 3G.4-5X to 4-10Z. The key locations are defined in Table 3G.4-1. The design seismic response spectra are servatively adjusted in the low frequency range in anticipation of future sites having a slightly er response at the lower frequency.
in-structure response spectra at six key locations, as defined below, are used if a site-specific 3D amic analysis evaluation as outlined in Subsection 2.5.2 is required. The site is acceptable if the r response spectra from the site-specific evaluation do not exceed the AP1000 spectra for each of locations identified below or the exceedances are justified.
[FRS Location Figure Numbers Containment internal structures at elevation of Figure 3G.4-5X to 3G.4-5Z reactor vessel support Containment operating floor Figure 3G.4-6X to 3G.4-6Z Auxiliary building NE corner at elevation 116'-6" Figure 3G.4-7X to 3G.4-7Z Shield building at fuel building roof Figure 3G.4-8X to 3G.4-8Z Staff approval is required prior to implementing a change in this information.
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Note:
See Table 3G.4-1 for locations of six key locations.
4.3.4 Bearing Pressure Demand ring pressure demand was calculated using both 2D and 3D analyses. Both linear and non-linear lyses are performed with the 2D nuclear island model. The maximum bearing pressures ulated include the effect of dead, live, and seismic loading.
2D model was used to evaluate the effect of liftoff on the bearing pressure. Since the largest ring pressure will result from the east-west seismic excitation because of the smaller width of the emat in this direction, liftoff was evaluated using an east-west stick model of the nuclear island ctures, supported on a rigid basemat with non-linear springs. Direct integration time history lyses were performed. The bearing pressures calculated from these analyses are summarized in le 3G.4-2. The pressures are at the edge of the basemat. Results are given for the three cases result in the highest bearing pressure (hard rock [HR], upper bound soft to medium [UBSM] soil, soft to medium [SM] soil). The linear results show maximum bearing pressures on the west side 1 to 33 ksf. Liftoff increases the subgrade pressure close to the west edge by 4 percent to rcent with insignificant effect beneath most of the basemat.
SASSI soil-structure interaction analyses are performed based on the nuclear island 3D SASSI el for the hard rock and five soil conditions established from the AP1000 2D SASSI analyses.
SASSI model of the nuclear island is based on the NI20 finite element model. The bearing sures from the 3D SASSI analyses have been obtained by combining the time history results the north-south, east-west, and vertical earthquakes. The maximum soil-bearing pressure and is obtained from the hard rock (HR) case equal to 35 ksf. It is noted that a maximum lized peak is obtained on the west edge of 38 ksf; a limit of 35 ksf for maximum bearing seismic and is obtained by averaging the soil pressure over 335 ft2 of the west edge of the shield building re the maximum stress occurs.
5 References NUREG-800, Review of Safety Analysis Reports for Nuclear Power Plants, Section 3.7.2, Seismic System Analysis, Revision 2.
GW-GL-700, AP600 Design Control Document, Appendices 2A and 2B, Revision 4.
APP-GW-S2R-010, Extension of Nuclear Island Seismic Analyses to Soil Sites, Westinghouse Electric Company LLC.
APP-GW-GLN-112, Structural Verification for Enhanced Shield Building, Westinghouse Electric Company LLC.
U.S. NRC Regulatory 1.92, Revision 2, Combining Modal Responses and Spatial Components in Seismic Analysis.
APP-GW-GLR-044, Nuclear Island Basemat and Foundation, Westinghouse Electric Company LLC.
Staff approval is required prior to implementing a change in this information.
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Analysis Type of Dynamic Model Method Program Response/Purpose (ASB) solid-shell - ANSYS Creates the finite element mesh for the ASB finite el element model.
(CIS) solid-shell - ANSYS Creates the finite element mesh for the CIS finite el element model.
inite element model - ANSYS ASB portion of NI10.
uding shield building (ASB10) inite element model Response spectrum ANSYS CIS portion of NI10.
uding dish below analysis tainment vessel finite element shell Mode superposition ANSYS Performed for hard rock profile for ASB with CIS el of nuclear island time history analysis as superelement and for CIS with ASB as 0] (coupled superelement.
iliary and shield ding shell model, To develop time histories for generating plant tainment internal design floor response spectra for nuclear island ctures, steel structures.
tainment vessel, r crane, RCL, To obtain maximum absolute nodal accelerations ssurizer, and CMTs) (ZPA) to be used in equivalent static analyses.
To obtain maximum displacements relative to basemat.
finite element Mode superposition ANSYS Performed for hard rock profile for comparisons rse shell model of time history analysis against more detailed NI10 model.
iliary and shield ding and tainment internal ctures [NI20]
luding steel tainment vessel, r crane, RCL, and ssurizer) 3G-11 Revision 1
Analysis Type of Dynamic Model Method Program Response/Purpose te element Time history analysis SASSI Performed 2D parametric soil studies to help ped-mass stick establish the bounding generic soil conditions and el of nuclear island to develop adjustment factors to reflect all generic site conditions for seismic stability evaluation.
te element Direct integration time ANSYS Performed 2D linear and non-linear seismic ped-mass stick history analysis analyses to evaluate effect of liftoff on Floor el of nuclear island Response Spectra and bearing.
finite element Time history analysis SASSI Performed for the five soil profiles of firm rock, soft rse shell model of Complex frequency rock, upper bound soft-to-medium soil, soft-to-iliary and shield response analysis medium soil, and soft soil.
ding and tainment internal To develop time histories for generating plant ctures [NI20] design floor response spectra for nuclear island luding steel structures.
tainment vessel, r crane, RCL, and To obtain maximum absolute nodal accelerations ssurizer) (ZPA) to be used in equivalent static analyses.
To obtain maximum displacements relative to basemat.
To obtain SSE bearing pressures for all generic soil cases.
To obtain maximum member forces and moments in selected elements for comparison to equivalent static results.
shell model of Mode superposition ANSYS Performed to develop loads for seismic stability iliary and shield time history analysis evaluation.
ding and tainment internal ctures [NI20]
luding steel tainment vessel) 3G-12 Revision 1
Analysis Type of Dynamic Model Method Program Response/Purpose shell of revolution Modal analysis; ANSYS To obtain dynamic properties.
el of steel equivalent static tainment vessel analysis using To obtain SSE stresses for the containment accelerations from vessel.
time history analyses umped-mass stick - ANSYS Used in the NI10 and NI20 models.
el of the SCV umped-mass stick - ANSYS Used in the NI10 and NI20 models.
el of the RCL umped-mass stick - ANSYS Used in the NI10 and NI20 models.
el of the ssurizer umped-mass stick - ANSYS Used in the NI10 model.
el of the CMT umped mass Modal analysis ANSYS To obtain dynamic properties.
iled model of the r crane Used with 3D finite element shell model of the containment vessel.
umped mass - ANSYS Used in the NI10 and NI20 models.
plified (single beam) el of the polar e
finite element shell Mode superposition ANSYS Used with detailed polar crane model to obtain el of containment time history analysis acceleration response of equipment hatch and sel airlocks.
Equivalent static To obtain shell stresses in vicinity of the large analysis penetrations of the containment vessel.
3G-13 Revision 1
Analysis Type of Dynamic Model Method Program Response/Purpose nite element refined Equivalent static non-linear ANSYS To obtain SSE member model of nuclear analysis using accelerations forces for the nuclear island d (NI05) from time history analyses basemat.
Mode superposition time To obtain floor and wall history analysis for the wall flexibility response and floor flexibility using characteristics.
synthetic time histories developed to match spectral To obtain maximum envelopes applied at the displacements relative to base basemat.
Response spectrum To obtain SSE member analysis with seismic input forces for the auxiliary and enveloping all soils cases shield building and the containment internal structures.
nite element coarse Mode superposition time ANSYS To obtain total basemat model of auxiliary and history analysis with seismic reactions for comparison to d building and input enveloping all soil reactions in equivalent static ainment internal cases linear analyses using NI05 tures [NI20] (including model.
l containment vessel, r crane, RCL, and surizer) drant model of shield Equivalent static analysis ANSYS To obtain member forces for ing roof (See shield building roof and section 3.8.4.4.1 for The PCS tank is designed radial roof beams, air inlet mation on use of the using the maximum structure, tension ring, and rant model.) accelerations at the PCS tank.
applicable elevation resulting from time history dynamic analyses of the nuclear island.
The tension ring and air inlet use maximum accelerations that are increased based on results of response spectrum analysis.
3G-14 Revision 1
Three Analysis Components Modal Model Method Program Combination Combination finite element, fixed Mode superposition time ANSYS Algebraic Sum n/a e models, coupled history analysis iliary and shield building ll model, with erelement of tainment internal ctures (NI10 and NI20) finite element nuclear Complex frequency SASSI Algebraic Sum n/a nd model (NI20) response analysis finite element, fixed Response spectrum ANSYS SRSS or Lindley-Yow e models, coupled analysis 100%, 40%, 40%
iliary and shield building containment internal ctures including shield ding roof (NI05) finite element model of Equivalent static analysis ANSYS 100%, 40%, 40% n/a nuclear island basemat using nodal accelerations
- 5) from shell model shell of revolution Equivalent static analysis ANSYS SRSS or n/a el of steel containment using nodal accelerations 100%, 40%, 40%
sel from 3D stick model S valve room and Response spectrum ANSYS SRSS or Grouping or cellaneous steel frame analysis 100%, 40%, 40% Lindley-Yow ctures, miscellaneous ble walls, and floors stick model analyses Direct integration time ANSYS Algebraic Sum n/a liftoff history 3G-15 Revision 1
Vessel Lumped-Mass Stick Model (Without Polar Crane) Modal Properties Effective Mass Mode Frequency X Direction Y Direction Z Direction 1 6.309 2.380 159.153 0.005 2 6.311 159.290 2.382 0.000 3 12.942 0.018 0.000 0.000 4 16.970 0.000 0.006 171.030 5 18.960 0.102 40.263 0.002 6 18.970 40.161 0.102 0.000 7 28.201 0.000 0.000 28.073 8 31.898 0.054 2.636 0.000 9 31.999 2.789 0.057 0.000 10 37.990 0.909 0.007 0.000 11 38.634 0.022 4.846 0.009 12 38.877 3.758 0.014 0.000 13 47.387 0.000 0.000 5.066 14 54.039 4.649 0.633 0.000 15 54.065 0.624 4.693 0.002 16 60.628 0.002 0.042 3.389 17 62.734 0.147 0.001 0.018 18 63.180 0.000 0.050 7.069 19 63.613 0.002 0.001 0.003 20 65.994 0.022 0.659 0.041 Sum of Effective Masses 214.929 215.545 214.706 s:
Fixed at Elevation 100.
The total mass of the containment vessel is 225.697 kip-sec2/ft.
3G-16 Revision 1
Modal Properties Effective Mass Mode Frequency X Direction Y Direction Z Direction 1 3.619 0.000 41.959 0.000 2 5.387 175.274 0.000 0.175 3 6.192 0.000 148.385 0.005 4 6.415 3.321 0.000 24.074 5 9.422 0.002 1.017 0.000 6 9.674 10.510 0.000 0.532 7 12.811 0.015 0.001 0.000 8 15.757 0.004 0.320 0.010 9 16.367 3.103 0.003 159.153 10 17.495 28.537 0.001 19.546 11 18.944 0.000 40.053 0.001 12 21.043 10.724 0.000 0.426 13 22.102 0.000 0.005 0.000 14 27.340 0.054 0.000 18.661 15 30.387 2.978 0.001 1.559 16 31.577 0.002 3.526 0.004 17 35.033 0.194 0.006 3.895 18 35.535 0.211 0.027 0.399 19 35.646 0.000 1.451 0.019 20 37.599 0.325 0.426 0.007 Sum of Effective Masses 235.254 237.181 228.465 s:
Fixed at Elevation 100.
The total mass of the containment vessel with the polar crane is 255.85 kip-sec2/ft.
3G-17 Revision 1
Vertical Model Horizontal Model Shell of Revolution Shell of Revolution Mode No. Model Stick Model Model Stick Model 1 16.51 hertz 16.97 hertz 6.20 hertz 6.31 hertz 2 23.26 hertz 28.20 hertz 18.58 hertz 18.96 hertz Fixed at elevation 100.
3G-18 Revision 1
North-South Hard Firm Soft Soft Rock Rock Rock UBSM SM Soil Node El. feet ZPA [g] ZPA [g] ZPA [g] ZPA [g] ZPA [g] ZPA [g]
ASB 21 81.5 0.326 0.326 0.345 0.358 0.306 0.249 41 99 0.348 0.327 0.347 0.361 0.308 0.227 120 179.6 0.571 0.501 0.469 0.498 0.529 0.247 150 242.5 0.803 0.795 0.816 0.819 0.787 0.29 310 333.1 1.449 1.561 1.567 1.524 1.226 0.453 CV 407 138.6 0.405 0.424 0.408 0.387 0.407 0.232 411 200 0.82 0.916 0.672 0.541 0.484 0.263 417 281.9 1.396 1.465 1.031 0.723 0.598 0.372 CIS 535 134.3 0.548 0.45 0.347 0.368 0.355 0.229 538 169 1.517 0.874 0.45 0.441 0.397 0.317 East-West Hard Firm Soft Soft Rock Rock Rock UBSM SM Soil Node El. feet ZPA [g] ZPA [g] ZPA [g] ZPA [g] ZPA [g] ZPA [g]
ASB 21 81.5 0.309 0.318 0.359 0.376 0.311 0.235 41 99 0.318 0.336 0.367 0.385 0.317 0.237 120 179.6 0.607 0.561 0.546 0.549 0.605 0.295 150 242.5 0.84 0.823 0.854 0.912 0.962 0.557 310 333.1 1.449 1.536 1.624 1.74 1.506 0.891 CV 407 138.6 0.528 0.529 0.535 0.513 0.38 0.247 411 200 0.817 0.95 0.816 0.741 0.515 0.429 417 281.9 1.251 1.503 1.136 0.985 0.716 0.675 CIS 535 134.3 0.52 0.404 0.391 0.404 0.365 0.259 538 169 1.679 1.052 0.755 0.553 0.526 0.441 3G-19 Revision 1
Location General Area Elevation (feet) at Reactor Vessel Support Elevation SCV Center 100.00 at Operating Deck SG West Compartment, NE 134.25 B NE Corner at Control Room Floor NE Corner 116.50 B Corner of Fuel Building Roof at Shield Building NW Corner of Fuel Bldg 179.19 B Shield Building Roof Area South Side of Shield Bldg 327.41 V Near Polar Crane SCV Stick Model 224.00 3G-20 Revision 1
East Edge West Edge Soil Case Analysis (ksf) (ksf) rd Rock Linear 17.18 32.77 Liftoff 17.38 34.85 per-bound Linear 19.46 31.69 ft to Medium Liftoff 18.42 33.51 ft to Medium Linear 15.84 30.82 Liftoff 17.06 32.18 3G-21 Revision 1
Figure 3G.1-1 Nuclear Island Seismic Analysis Models 3G-22 Revision 1
Figure 3G.2-1 3D Finite Element Model of Coupled Shield and Auxiliary Building 3G-23 Revision 1
Figure 3G.2-2 3D Finite Element Model of Containment Internal Structures 3G-24 Revision 1
Figure 3G.2-3 3D Finite Element Model of Containment Outer Basemat (Dish) 3G-25 Revision 1
Figure 3G.2-4 Steel Containment Vessel and Polar Crane Models 3G-26 Revision 1
Figure 3G.2-5A Polar Crane Model Simplified Model 3G-27 Revision 1
Figure 3G.2-5B Polar Crane Model Detailed Model 3G-28 Revision 1
Figure 3G.2-6 Reactor Coolant Loop Lumped-Mass Stick Model 3G-29 Revision 1
Figure 3G.2-7 Pressurizer Model 3G-30 Revision 1
Figure 3G.2-8 Core Makeup Tank Models 3G-31 Revision 1
Figure 3G.2-9 AP1000 Nuclear Island Solid-Shell Model (NI10) 3G-32 Revision 1
Figure 3G.2-10 ntainment Internal Structure with the SCV, PC, Reactor Coolant Loop, and Pressurizer 3G-33 Revision 1
Figure 3G.2-11 Soil Structure Interaction Model - NI20 Looking East 3G-34 Revision 1
Figure 3G.2-12 Coarse Model of Containment Internal Structures 3G-35 Revision 1
Figure 3G.2-13 Fine Mesh (NI05) Model of Auxiliary and Shield Building 3G-36 Revision 1
Figure 3G.2-14 NI05 Model of Containment Internal Structures 3G-37 Revision 1
Figure 3G.2-15 3D NI05 Refined Mesh Model of Outer Containment Basemat (Dish) 3G-38 Revision 1
Figure 3G.2-16 Quadrant Model of Shield Building Roof 3G-39 Revision 1
Figure 3G.2-17 Detailed 3D Finite Element Model of Containment Vessel Including Large Penetrations 3G-40 Revision 1
Figure 3G.2-18 Axisymmetric Model of Containment Vessel 3G-41 Revision 1
Figure 3G.2-19 Schematic of Non-linear 2D East-West Nuclear Island Stick Model Used for Stability Evaluation that Addresses Sliding and Overturning 3G-42 Revision 1
Figure 3G.3-1 Generic Soil Profiles 3G-43 Revision 1
Figure 3G.3-2 2D SASSI FRS - Node 41 X (ASB El. 99) 3G-44 Revision 1
Figure 3G.3-3 2D SASSI FRS - Node 41 Y (ASB El. 99) 3G-45 Revision 1
Figure 3G.3-4 2D SASSI FRS - Node 120 X (ASB El. 179.6) 3G-46 Revision 1
Figure 3G.3-5 2D SASSI FRS - Node 120 Y (ASB El. 179.6) 3G-47 Revision 1
Figure 3G.3-6 2D SASSI FRS - Node 310 X (ASB El. 333.2) 3G-48 Revision 1
Figure 3G.3-7 2D SASSI FRS - Node 310 Y (ASB El. 333.2) 3G-49 Revision 1
Figure 3G.3-8 2D SASSI FRS - Node 411 X (SCV El. 200.0) 3G-50 Revision 1
Figure 3G.3-9 2D SASSI FRS - Node 411 Y (SCV El. 200.0) 3G-51 Revision 1
Figure 3G.3-10 2D SASSI FRS - Node 535 X (CIS El. 134.3) 3G-52 Revision 1
Figure 3G.3-11 2D SASSI FRS - Node 535 Y (CIS El. 134.3) 3G-53 Revision 1
Figure 3G.4-1 Auxiliary Shield Building Rigid Nodes at El. 135 3G-54 Revision 1
Figure 3G.4-2 Auxiliary Shield Building Flexible Nodes at El. 135 3G-55 Revision 1
Figure 3G.4-3 Excavated Soil 3G-56 Revision 1
Figure 3G.4-4 Additional Elements for Soil Pressure Calculations 3G-57 Revision 1
[ Figure 3G.4-5X X Direction FRS for Node 130401 (NI10) or 1761 (NI20) CIS at Reactor Vessel Support Elevation of 100]*
Staff approval is required prior to implementing a change in this information.
3G-58 Revision 1
[Figure 3G.4-5Y Y Direction FRS for Node 130401 (NI10) or 1761 (NI20) CIS at Reactor Vessel Support Elevation of 100]*
Staff approval is required prior to implementing a change in this information.
3G-59 Revision 1
[Figure 3G.4-5Z Z Direction FRS for Node 130401 (NI10) or 1761 (NI20) CIS at Reactor Vessel Support Elevation of 100]*
Staff approval is required prior to implementing a change in this information.
3G-60 Revision 1
[Figure 3G.4-6X X Direction FRS for Node 105772 (NI10) or 2199 (NI20) CIS at Operating Deck Elevation 134.25]*
Staff approval is required prior to implementing a change in this information.
3G-61 Revision 1
[Figure 3G.4-6Y Y Direction FRS for Node 105772 (NI10) or 2199 (NI20) CIS at Operating Deck Elevation 134.25]*
Staff approval is required prior to implementing a change in this information.
3G-62 Revision 1
[Figure 3G.4-6Z Z Direction FRS for Node 105772 (NI10) or 2199 (NI20) CIS at Operating Deck Elevation 134.25]*
Staff approval is required prior to implementing a change in this information.
3G-63 Revision 1
[Figure 3G.4-7X X Direction FRS for Node 4724 (NI10) or 2078 (NI20) ASB Control Room Side Elevation 116.50]*
Staff approval is required prior to implementing a change in this information.
3G-64 Revision 1
[Figure 3G.4-7Y Y Direction FRS for Node 4724 (NI10) or 2078 (NI20) ASB Control Room Side Elevation 116.50]*
Staff approval is required prior to implementing a change in this information.
3G-65 Revision 1
[Figure 3G.4-7Z Z Direction FRS for Node 4724 (NI10) or 2078 (NI20) ASB Control Room Side Elevation 116.50]*
Staff approval is required prior to implementing a change in this information.
3G-66 Revision 1
[Figure 3G.4-8X X Direction FRS for Node 5754 (NI10) or 2675 (NI20) ASB Fuel Building Roof Elevation 179.19]*
Staff approval is required prior to implementing a change in this information.
3G-67 Revision 1
[Figure 3G.4-8Y Y Direction FRS for Node 5754 (NI10) or 2675 (NI20) ASB Fuel Building Roof Elevation 179.19]*
Staff approval is required prior to implementing a change in this information.
3G-68 Revision 1
[Figure 3G.4-8Z Z Direction FRS for Node 5754 (NI10) or 2675 (NI20) ASB Fuel Building Roof Elevation 179.19]*
Staff approval is required prior to implementing a change in this information.
3G-69 Revision 1
[Figure 3G.4-9X X Direction FRS for Node 8573 (NI10) or 3329 (NI20) ASB Shield Building Roof Elevation 327.41]*
Staff approval is required prior to implementing a change in this information.
3G-70 Revision 1
[Figure 3G.4-9Y Y Direction FRS for Node 8573 (NI10) or 3329 (NI20) ASB Shield Building Roof Elevation 327.41]*
Staff approval is required prior to implementing a change in this information.
3G-71 Revision 1
[Figure 3G.4-9Z Z Direction FRS for Node 8573 (NI10) or 3329 (NI20) ASB Shield Building Roof Elevation 327.41]*
Staff approval is required prior to implementing a change in this information.
3G-72 Revision 1
[Figure 3G.4-10X X Direction FRS for Node 130412 (NI10) or 2788 (NI20) SCV Near Polar Crane Elevation 224.00]*
Staff approval is required prior to implementing a change in this information.
3G-73 Revision 1
[Figure 3G.4-10Y Y Direction FRS for Node 130412 (NI10) or 2788 (NI20) SCV Near Polar Crane Elevation 224.00]*
Staff approval is required prior to implementing a change in this information.
3G-74 Revision 1
[ Figure 3G.4-10Z Z Direction FRS for Node 130412 (NI10) or 2788 (NI20) SCV Near Polar Crane Elevation 224.00]*
Staff approval is required prior to implementing a change in this information.
3G-75 Revision 1
s appendix summarizes the structural design and analysis of structures identified as "Critical tions" in the auxiliary and shield buildings. The design summaries include the following rmation:
Description of buildings Governing codes and regulations Structural loads and load combinations Global analyses Structural design of critical structural elements sections 3H.2 through 3H.5 include a general description of the auxiliary building and shield ding, a summary of the design criteria and the global analyses. Examples of the structural design shown for 14 critical sections which are identified in subsection 3H.5 and shown in res 3H.5-1 (3 sheets). The exact locations of the critical sections related to the shield building nder shown in Figure 3H.5-16. Representative design details are provided for these structures in section 3H.5.]*
2 Description of Auxiliary and Shield Buildings 2.1 Description of Auxiliary Building e auxiliary building is a reinforced concrete structure. The auxiliary building is one of the three dings that make up the nuclear island and shares a common basemat with the containment ding and the shield building. The auxiliary building general layout is shown in Figure 3H.2-1. It is a haped section of the nuclear island that wraps around approximately half of the circumference of shield building. The building dimensions are shown on key structural dimension drawings, re 3.7.2-12.
auxiliary building is divided into six areas, which are identified in Figure 3H.2-1. It is a 5-story ding; three stories are located above grade and two are located below grade. Areas 1 and 2 ure 3H.2-1) have five floors, including two floors below grade level. The lowest floor at ation 66-6 is used exclusively for housing battery racks. The next higher floor, at ation 82-6, also has battery racks and some electrical equipment. The floor at the grade level, ation 100-0, has electrical penetration areas, a remote shutdown workstation room, and some sion A and Division C equipment. The main control room is situated on the floor at ation 117-6, which also has rooms for the main steam and feedwater lines. The floor at ation 135-3 carries air filtration and air handling units, chiller pumps, and other mechanical and trical equipment. The roof for areas 1 and 2 is at elevation 153-0.
as 3 and 4 of the auxiliary building are the areas east of the containment shield building. Valve piping areas, and some mechanical equipment, are located in the basement floor at ation 66-6. The floor at elevation 82-6 has a piping penetration area, a radiation chemistry ratory, makeup pumps, and other mechanical equipment. The floor at grade level ation 100-0 has an electrical penetration room, a staging area for the equipment hatch, and the ess opening to the annex building. The electrical penetration area, trip switchgears, and motor trol centers occupy most of the floor at elevation 117-6. The floor at elevation 135-3 is used for Staff approval is required prior to implementing a change in this information.
3H-1 Revision 1
as 5 and 6 include facilities for storage and handling of new and spent fuel. The spent fuel pool, transfer canal, and cask loading and cask washdown pits have concrete walls and floors. They lined on the inside surface with stainless steel plate for leak prevention. The walls and major rs are constructed using concrete filled steel plate modules. The new fuel storage area is a arate reinforced concrete pit providing temporary dry storage for the new fuel assemblies. A
-ton cask handling crane travels in the east-west direction. The location and travel of this crane ents the crane from carrying loads over the spent fuel pool to preclude them from falling into the nt fuel pool. Mechanical equipment is also located in this area for spent fuel cooling, residual heat oval, and liquid waste processing. This equipment is generally nonsafety-related.]*
2.2 Description of Shield Building shield building is the structure and annulus area that surrounds the containment building. It res a common basemat with the containment building and the auxiliary building. The shield ding uses concrete-filled steel plate construction (SC) as well as reinforced concrete (RC) cture. The figures in Section 1.2 show the layout of the shield building and its interface with the r buildings of the nuclear island.
re 3.8.4-5 shows the following significant features and the principal systems and components of shield building:
Shield building cylindrical structure Shield building roof structure RC/SC connections Air inlets and tension ring Knuckle region (connection to exterior wall of PCS tank)
Compression ring (connection to interior wall of PCS tank)
Passive containment cooling system (PCS) water storage tank (PCCWST) overall configuration of the shield building is established from functional requirements related to ation shielding, missile barrier, passive containment cooling, tornado, and seismic event ection. These functional requirements led to establishing the design based on two primary design es used for nuclear plant structures: 1) ACI 349 for reinforced concrete design, and NSI/AISC N690 for structural steel design.
shield building SC walls are anchored to the RC basemat and shield building RC wall by hanical connections. These RC-to-SC connections are also used in the other regions of the ld building, including:
Auxiliary building RC roof connection to the shield building SC wall Auxiliary building RC wall connection to shield building SC wall Tension ring connection to the shield building RC roof connections provide for the direct transfer of forces from the RC reinforcing steel to the SC liner es.
cylindrical shield wall has an outside radius of 72.5 feet and a thickness of 36 inches. The ndrical wall section that is a few feet below the auxiliary building roof line is a reinforced concrete
) structure. The section that is not protected by the auxiliary building is a steel concrete (SC) posite structure (see Figure 3H.5-16). The overall thickness of 36 inches is the same as the RC below. The concrete for the SC portion is standard concrete with compressive strength of Staff approval is required prior to implementing a change in this information.
3H-2 Revision 1
l plate. The tie bar spacing is reduced in the higher stress regions. A typical SC wall panel is wn in Figure 3H.5-13.
tension ring is located at the interface of the shield building steel concrete composite air inlet ctures and the shield building reinforced concrete roof. The top of the tension ring interfaces with RC roof slab. The tension ring supports the roof girders that are located under the RC roof slab.
bottom of the tension ring is attached to the air inlets structure. The bottom of the air inlets cture is attached to the top of the cylindrical SC wall of the shield building. The connection of the ion ring to the roof is of RC design and is described above.
primary function of the tension ring is to resist the thrust from the shield building roof. The air s structure is located directly below the tension ring and includes the air openings that provide for ral circulation of cooling air. Though its steel plates are connected to the concrete infill by studs tie bars, the tension ring is conservatively designed as a hollow steel box girder. The concrete is credited only for out-of-plane shear transfer and for stability of the steel plates. The tension is designed to have high stiffness and to remain elastic under required load combinations.
air inlets structure is a 4.5-foot-thick SC structure with through-wall openings for air flow. The air openings consist of circular pipes at a downward inclination of 38 degrees from the vertical.
el plates on each face, aligned with the inner and outer flanges of the tension ring, serve as ary reinforcement. The concrete infill is connected to the steel plates with tie bars and studs. The of the air inlets structure is welded to the underside of the tension ring. The bottom of the air inlets cture is welded to the SC wall.
shield building conical roof steel structure consists of 32 radial beams. Between each pair of al beams there are circumferential beams. A steel plate is welded to the top flanges of each beam forms a surface on which the concrete is placed. The steel structure forms a conical shell that ns the area from the compression ring to the tension ring.
outside diameter of the PCS tank (passive containment cooling water storage tank) intersects the shield building roof at the knuckle region. Outside of the PCS tank, the concrete roof slab kness is 3 feet and at the bottom of the PCS tank, the concrete thickness is 2 feet. The wall from PCS tank applies a load to the roof slab, and also provides stiffness and increases the strength of roof in that region.
inside diameter of the PCS tank intersects with the roof slab at the compression ring. The pression ring provides the compression support for the conical roof dome. It consists of a posite structure having a curved steel beam section, which supports the concrete roof directly ve it. The inside wall of the PCS tank is located above the concrete roof. Studs are placed on the flange of the steel girder to allow the steel and concrete sections to act as a composite unit. The ed girder is designed to provide support for the steel structure during construction and during the al placement of the concrete roof before the concrete has hardened sufficiently.
PCS tank sits on top of the shield building roof. It is supported by and acts integrally with the ical roof. The inside surface has a liner that functions to provide leak protection, but is not uired to provide structural strength to the structure. Leak chase channels are provided over the r welds. The top elevation of the water inside the tank for the PCS has sufficient freeboard to lude impact on the roof during the SSE.
3H-3 Revision 1
horizontal concrete slabs supported by composite structural steel framing.
Seismic forces are obtained from the response spectrum analysis of the three-dimensional finite element analysis models as described in subsection 3H.4. The shear wall and floor slab design also considers out-of-plane bending and shear forces due to loading, such as live load, dead load, seismic, lateral earth pressure, hydrostatic, hydrodynamic, and wind pressure.
The shield building roof and the passive containment cooling water storage tank are analyzed using three-dimensional finite element models with the ANSYS computer code]* as described in subsection 3.8.4.4.1. [Loads and load combinations include construction, dead, live, thermal, wind, and seismic. The response spectrum analysis of the nuclear island is supplemented by equivalent static acceleration analysis of a more detailed model of a quadrant of the shield building roof. The results from the more detailed analysis are used in the evaluation of the tension ring, air inlets, and radial beams. The seismic response of the water in the tank is analyzed in a separate analysis with seismic input defined by the floor response spectrum.
The structural steel framing is used primarily to support the concrete slabs and roofs. Metal decking, supported by the steel framing, is used as form work for the concrete slabs and roofs.
The finned floors for the main control room and the instrumentation and control room ceilings are designed as reinforced concrete slabs in accordance with American Concrete Institute standard ACI 349. The steel panels are designed and constructed in accordance with American Institute of Steel Construction Standard AISC N690. For positive bending, the steel plate is in tension and the steel plate with fin stiffeners serves as the bottom reinforcement.
For negative bending, compression is resisted by the stiffened plate and tension by top reinforcement in the concrete.]*
3.1 Governing Codes and Standards e primary codes and standards used in the design of the auxiliary and shield buildings are listed w:
ACI 349-01, "Code Requirement for Nuclear Safety-Related Structure Steel" (refer to subsection 3.8.4.5 for supplementary requirements)
ANSI/AISC N690-1994, "Specification for the Design, Fabrication and Erection of Safety-Related Steel Structures for Nuclear Facilities" (refer to subsection 3.8.4.5 for supplemental requirements).]*
3.2 Seismic Input SSE design response spectra are given in Figures 3.7.1-1 and 3.7.1-2. [They are based on the ulatory Guide 1.60 response spectra anchored to 0.30g, but are amplified at 25 Hertz to reflect er high-frequency seismic energy content observed for eastern United States sites.]* The nuclear nd seismic analyses are summarized in Subsection 3.7.2.
Staff approval is required prior to implementing a change in this information.
3H-4 Revision 1
sections are used for the design of the building structures. All the listed loads are not necessarily licable to all structures and their elements. Loads for which each structural element is designed based on the conditions to which that particular structural element is potentially subjected.]*
d Load (D):
e weight of all permanent construction and installations, including fixed equipment, is included as dead load during its normal operating condition.
weight of minor equipment (not specifically included in the dead load), piping, cables and cable s, ducts, and their supports was included as equivalent dead load (EDL). A minimum of ounds per square foot (psf) was used as EDL. For floors with a significant number of small es of equipment, the total weight of miscellaneous small pieces of equipment, divided by the floor a of the room plus an additional 50 psf was used as the equivalent dead load.]*
th Pressure (H):
e static earth pressure acting on the structures during normal operation is considered in the ign of exterior walls. The dynamic soil pressure, induced during a safe shutdown earthquake E), is included as a seismic load.]*
e Loads (L):
e load imposed by the use and occupancy of the building is included as the live load. Live loads ude floor area loads, laydown loads, fuel transfer casks, equipment handling loads, trucks, oad vehicles, and similar items. The floor area live load is not applied on areas occupied by ipment whose weight is specifically included in the dead load. Live load is applicable on areas er equipment where access is provided, for instance, the floor under an elevated tank supported egs.
r loading diagrams are prepared for areas for component laydown. The diagrams show the tion of major pieces of equipment and their foot-print loads or equivalent uniformly distributed s.
following live load items are considered in design:
Building floor loads following minimum values for live loads are used.
- Structural platforms and gratings 100 psf
- Ground floors 250 psf
- All other elevated floors 200 psf (This load is reduced if the equivalent dead load for the floor is more than 50 psf. The sum of the live load and the equivalent dead load is 250 psf.)
Staff approval is required prior to implementing a change in this information.
3H-5 Revision 1
corresponds to ground snow load of 75 psf, exposure factor of 1.0, thermal factor of 1.0, and an ortance factor of 1.2.
Concentrated loads for the design of local members
- Concentrated load on beams and 5,000 pounds so applied as to maximize girders (in load combinations that moment or shear. This load is not carried to do not include seismic load) columns or walls. It is not applied in areas where no heavy equipment will be located or transported, such as the access control areas.
- Concentrated load on slabs 5,000 pounds so applied as to maximize (considered with dead load only) moment or shear. This load is not carried to columns or walls. It is not applied in access control areas.
esign reconciliation analysis, if actual loads are established to be lower than the above loads, the al loads are used for reconciliation.
Temporary exterior wall surcharge en applicable, a minimum surcharge outside and adjacent to subsurface wall of 250 psf is applied.
Construction loads additional construction loads produced by cranes, trucks, and the like, with their pickup loads, considered. For steel beams supporting concrete floors, the weight of the wet concrete plus psf uniform load and 5,000 pounds concentrated load, distributed near points of maximum shear moment, is applied. A one-third increase in allowable stress is permitted.
al decking and precast concrete panels, used as formwork for concrete floors are designed for wet weight of the concrete plus a construction live load of 20 psf uniform or 150 pounds centrated. The deflection during normal operation is limited to span in inches divided by 180, or inch, whichever is less.
Crane loads impact allowance for traveling crane supports and runway horizontal forces is in accordance with C N690.
Elevator loads impact allowance used for the elevator supports is 100 percent, applied to design capacity and ght of car plus appurtenances, unless otherwise specified by the equipment supplier.
Equipment laydown and major maintenance rs are designed for planned refueling and maintenance activities as defined on equipment own drawings.]*
Staff approval is required prior to implementing a change in this information.
3H-6 Revision 1
Design wind (W)
For the design of the exterior walls, wind loads are applied in accordance with ASCE 7-98 with a basic wind speed of 145 mph. The importance factor is 1.15, and the exposure category is C.
Wind loads are not combined with seismic loads.
Tornado load (Wt)
The exterior walls of the auxiliary and shield buildings are designed for tornado. A maximum wind speed of 300 mph (maximum rotational speed: 240 mph, maximum translational speed: 60 mph) is used to design the structures.]*
smic Loads (Es) e SSE (Es) is used for evaluation of the structures of the auxiliary and shield buildings. Es is ned as the loads generated by the SSE specified for the plant, including the associated rodynamic loads and dynamic incremental soil pressure.]*
erating Thermal Loads (To) rmal thermal loads for the exterior walls and roofs are addressed in the design. These correspond ositive and negative linear temperature gradients with the inside surface at an average 70°F and outside air temperature at -40°F and +115°F, respectively. These loads are considered for the mic Category I structures in combination with the SSE also. All exterior walls of the nuclear island ve grade not protected by adjacent buildings are designed for these thermal loads. The thermal dient is also applied to the portion of the shield building between the upper annulus and the iliary building.
mal thermal loads for the passive containment cooling system (PCS) tank design are calculated ed on the outside air temperature extremes specified for the safety-related design. The PCS tank ssumed to be at 40°F when the outside air temperature is -40°F. The water in the PCS tank is umed to be at 70°F when the outside air temperature is postulated to be at 115°F.
mal thermal loads due to a thermal gradient in the structures below the grade level (exterior walls basemat) are small and are not considered in the design.]*
cts of Pipe Rupture (Y) e evaluations consider the following loads:
Accident design pressure load, Pa, within or across a compartment and/or building generated by the postulated pipe rupture, including the dynamic effects due to the pressure time history.
Main steam isolation valve (MSIV) and steam generator blowdown valve compartments are designed for a pressurization load of 6 pounds per square inch (psi).
Accident thermal loads, Ta, due to thermal conditions generated by the postulated pipe break and including To.
perature gradients are based on an exterior air temperature of -40°F.
Staff approval is required prior to implementing a change in this information.
3H-7 Revision 1
3.4 Load Combinations and Acceptance Criteria ncrete structures are designed in accordance with ACI 349 for the load combinations and load ors given in Table 3.8.4-2. Steel structures are designed in accordance with AISC N690 for the combinations and stress limit coefficients given in Table 3.8.4-1. The following supplemental uirements are applied for the use of AISC N690:
In Section Q1.0.2, the definition of secondary stress applies to stresses developed by temperature loading only.
In Section Q1.3, where the structural effects of differential settlement are present, they are included with the dead load, D.
In Table Q1.5.7.1, the stress limit coefficients for compression are as follows:
- 1.3 instead of 1.5 in load combinations 2, 5, and 6
- 1.4 instead of 1.6 in load combinations 7, 8, and 9
- 1.6 instead of 1.7 in load combination 11 In Section Q1.5.8, for constrained members (rotation and/or displacement constraint such that a thermal load causes significant stresses) supporting safety-related structures, systems, or components, the stresses under load combinations 9, 10, and 11 are limited to those allowed in Table Q1.5.7.1 as modified above.]*
4 Seismic Analyses lobal seismic analysis of the AP1000 nuclear island structure is performed to obtain building mic response for the seismic design of nuclear safety-related structures. The seismic loads for design of the shear walls and the slabs in the auxiliary building are based on a response ctrum analysis of the auxiliary building and the shield building 3D finite element models.]* This lysis is described in Subsection 3.7.2. [For determining the out-of-plane seismic loads on flexible s and wall segments, spectral accelerations are obtained from time history analyses or from the vant response spectra, using the 7 percent damping curve. Hand calculations are performed to mate the out-of-plane seismic forces and the corresponding bending moment in each shear wall floor slab element to supplement the loads obtained from the global seismic analysis.]*
4.1 Live Load for Seismic Design or live loads, based on requirements during plant construction and maintenance activities, are cified varying from 50 to 250 pounds per square foot.
the local design of members, such as the floors and beams, seismic loads include the response to masses equal to 25 percent of the specified floor live loads or 75 percent of the roof snow
, whichever is applicable. These seismic loads are combined with 100 percent of the specified loads, or 75 percent of the roof snow load, whichever is applicable. These live and snow loads included as mass in calculating the vertical seismic forces on the floors and roof. The mass of ipment and distributed systems is included in both the dead and seismic loads.]*
Staff approval is required prior to implementing a change in this information.
3H-8 Revision 1
ments in the auxiliary building and shield building. These structures are listed below and the esponding location numbers are shown on Figure 3H.5-1. The basis for their selection to this list so provided for each structure.
South wall of auxiliary building (column line 1), elevation 66-6 to elevation 180-0. (This exterior wall illustrates typical loads such as soil pressure, surcharge, temperature gradients, seismic, and tornado.) - see subsection 3H.5.1.1 and Figures 3H.5-2 and 3H.5-3 Interior wall of auxiliary building (column line 7.3), elevation 66-6 to elevation 160-6 (This is one of the most highly stressed shear walls.) - see subsection 3H.5.1.2 and Figure 3H.5-4 West wall of main control room in auxiliary building (column line L), elevation 117-6 to elevation 153-0. (This illustrates design of a wall for subcompartment pressurization.) -
see subsection 3H.5.1.3 and Figure 3H.5-12 North wall of MSIV east compartment (column line 11 between column lines L and M),
elevation 117-6 to elevation 153-0. (The main steam line is anchored to this wall segment.) - see subsection 3H.5.1.4 and Figure 3H.5-5 Roof slab at elevation 180-0 adjacent to shield building cylinder. (This is the connection between the two buildings at the highest elevation.) - see subsection 3H.5.2.1 and Figure 3H.5-7 Floor slab on metal decking at elevation 135-3. (This is a typical slab on metal decking and structural steel framing.) - see subsection 3H.5.2.2 and Figure 3H.5-6 2-0 slab in auxiliary building (operations work area (tagging room) ceiling) at elevation 135-3. (This illustrates the design of a typical 2-0 thick concrete slab.) - see subsection 3H.5.3.1 and Figure 3H.5-8. (Note: The Tagging Room has been renamed as Operations Work Area. However, to avoid changing the associated design and analysis documents, this room is referred to as the Tagging Room.)
Finned floor in the main control room at elevation 135-3. (This illustrates the design of the finned floors.) - see subsection 3H.5.4 and Figure 3H.5-9 Shield building roof/exterior wall of PCS water storage tank. (This is a unique area of the roof and water tank.) - see subsection 3H.5.6.3 Shield building roof/interior wall of PCS water storage tank. (This is a unique area of the roof and water tank.) - see subsection 3H.5.6.2 Shield building roof, tension ring, and air inlet. (This is the junction between the shield building roof and the cylindrical wall of the shield building.) - see subsections 3H.5.6 and 3H.5.6.1 Divider wall between the spent fuel pool and the fuel transfer canal. (This wall is subjected to thermal and seismic sloshing loads.) - see subsection 3H.5.5.1 and Figure 3H.5-10 Shield building SC cylinder is the exposed portions of the shield building that are not protected by the Auxiliary Building and is a steel concrete composite structure - see Staff approval is required prior to implementing a change in this information.
3H-9 Revision 1
Shield building SC to RC connection is the region of the shield building that anchors the SC cylindrical wall modules to the RC basemat and wall of the shield building - see subsection 3H.5.7.2, Figure 3H.5-16, and Figures 1, 2, and 3 of APP-GW-GLR-602 (Reference 1) design implemented in fabrication and construction drawings and instructions will have the ign shown, an equal design, or a better design for the key structural elements.]*
5.1 Shear Walls uctural Description ear walls in the auxiliary building vary in size, configuration, aspect ratio, and amount of forcement. The stress levels in shear walls depend on these parameters and the seismic eleration level. The range of these parameters and the stress levels in various regions of the most erely stressed shear wall are described in the following paragraphs.
height of the major structural shear walls in the auxiliary building ranges between 30 to 120 feet.
length ranges between 40 and 260 feet. The aspect ratio of these walls (full height/full length) is erally less than 1.0 and often less than 0.25. The walls are typically 2 to 5 feet thick, and are olithically cast with the concrete floor slabs, which are 9 inches to 2 feet thick. Exterior shear s are several stories high and do not have many large openings. Interior shear walls, however, discontinuous in both vertical and horizontal directions. The in-plane behavior of these shear s, including the large openings, is adequately represented in the analytical models for the global mic response. Where the refinement of these finite element models is insufficient for design of reinforcement, for example in walls with a large number of openings, detailed finite element els are used.
shear walls are used as the primary system for resisting the lateral loads, such as earthquakes.
auxiliary building shear walls are also evaluated for flexure and shear due to the out-of-plane s.]*
ign Approach e auxiliary building shear walls are designed to withstand the loads specified in section 3H.3.3. Beside dead, live, and other normal operating condition loads, the following loads considered in the shear wall design:
Seismic loads
- The SSE loads for the wall are obtained from the seismic analyses of auxiliary/shield buildings that are described in subsection 3H.4.
- Calculations are performed by considering shear wall segments bounded by the floors below and above the segment and the adjacent walls perpendicular to, on both sides of, the segment under consideration. Appropriate boundary conditions are assumed for the four edges of the segment. Natural frequencies of wall segments are determined using finite element models or text book formulas for the frequency of plate structures.
Corresponding spectral acceleration is determined from the applicable response spectrum.
Staff approval is required prior to implementing a change in this information.
3H-10 Revision 1
z Dynamic earth pressure calculated in accordance with ASCE 4-98 z Passive earth pressure Accident pressure load
- Shear walls of the main steam isolation valves (MSIV) rooms are designed for 6 pounds per square inch (psi) differential pressure acting in conjunction with the seismic loads.
Member forces due to accident pressure and SSE are combined by absolute sum.
- The main control room wall of the east MSIV compartment is evaluated for the pressure and the jet load due to a postulated main steamline break.
Tornado load For exterior walls above grade level, tornado loads are considered.
design temperatures for thermal gradient are included in Table 3H.5-1.
shear walls are designed for the load combinations, as applicable, contained in Table 3.8.4-2.
wall sections are designed in accordance with the requirements of ACI 349-01.]*
5.1.1 Exterior Wall at Column Line 1 e wall at column line 1 is the exterior wall at the south end of the nuclear island. The reinforced crete wall extends from the top of the basemat at elevation 66-6 to the roof at elevation 180-0.
3-0 thick below the grade and 2-3 thick above the grade.
wall is designed for the applicable loads including dead load, live load, hydrostatic load, static dynamic lateral soil pressure loads, seismic loads, and thermal loads. For various segments of wall, Table 3H.5-2 provides the listing and magnitude of the various design loads and le 3H.5-3 presents the details of the wall reinforcement. The sections where the required forcement is calculated are shown in Figure 3H.5-2 (Sheet 1). Typical wall reinforcement is wn on Figure 3H.5-3.]*
5.1.2 Wall at Column Line 7.3 e wall at column line 7.3 is a shear wall that connects the shield building and the nuclear island rior wall at column line I. It extends from the top of the basemat at elevation 66-6 to the top of roof. The wall is 3 feet thick below the grade at elevation 100-0 and 2 feet thick above the grade.
-of-plane lateral support is provided to the wall by the floor slabs on either side of it and the roof at top.
auxiliary building design loads are described in Section 3H.3.3, and the wall is designed for the licable loads.
various segments of this wall, the corresponding governing load combination and associated ign loads are shown in Table 3H.5-4. Table 3H.5-5 presents the details of the wall reinforcement.
sections where the required reinforcement is calculated are shown in Figure 3H.5-2 (Sheet 2).
cal wall reinforcement is shown on Figure 3H.5-4]*
Staff approval is required prior to implementing a change in this information.
3H-11 Revision 1
top of the basemat at elevation 66-6 to the top of the roof. The wall is 2 feet thick. Out-of-plane ral support is provided to the wall by the floor slabs on either side of it and the roof at the top. The ment of the wall that is a part of the main control room boundary is from elevation 117-6 to ation 135-3.
auxiliary building design loads are described in subsection 3H.3.3, and the wall is designed for applicable loads. In addition to the dead, live and seismic loads, the wall is designed to withstand pounds per square inch pressure load due to a pipe break in the MSIV room even though it is a ak exclusion area. This wall segment is also designed to withstand a jet load due to the pipe ak.
governing load combination and associated design loads are those due to the postulated pipe ure and are shown in Table 3H.5-6. Table 3H.5-7 and Figure 3H.5-12 present the details of the reinforcement. The sections where the required reinforcement is calculated are shown in re 3H.5-2 (Sheet 3).]*
5.1.4 Wall at Column Line 11 e north wall of the MSIV east compartment, at column line 11 between elevation 117-6 and ation 153-0, has been identified as a critical section.
segment of the wall between elevation 117-6 and elevation 135-3 is 4 feet thick, and several s such as the main steam line, main feed water line, and the start-up feed water line are hored to this wall at the interface with the turbine building.
wall segment from elevation 135-3 to elevation 153-0 does not provide support to any high rgy lines, and is 2 feet thick. This portion does not have to withstand reactions from high energy breaks.
wall is designed to withstand loads such as the dead load, live load, seismic load and the mal load. The MSIV room is a break exclusion area, but the design also considered the loads ociated with one square foot pipe rupture in the MSIV room, such as compartment pressurization, oad, and the reactions at the pipe anchors. The loads on the pipe anchor include pipe rupture s for breaks in the turbine building.
wall structure is analyzed using three dimensional finite element analyses supplemented by d calculations. Analyses are performed for individual loads, and design loads are determined for licable load combinations from Table 3.8.4-2.
cal wall reinforcement is shown in Figure 3H.5-5.]*
5.2 Composite Structures (Floors and Roof) e floors consist of a concrete slab on metal deck, which rests on structural steel floor beams.
eral floors in the auxiliary building are designed as one-way reinforced concrete slabs supported tinuously on steel beams. Typically, the beams span between two reinforced concrete walls. The ms are designed as composite with formed metal deck spanning perpendicular to the members.
hored construction is used. For the floors, beams are typically spaced at about 6-feet intervals spans are between 16 feet and 25 feet.]*
Staff approval is required prior to implementing a change in this information.
3H-12 Revision 1
ed perpendicular to each other. The depth of the ribs for 9-inch concrete floor slabs and 15-inch p concrete roof slabs are 3 inches and 4.5 inches respectively. The concrete slab is tied to the ctural steel floor beam by shear connectors, which are welded to the top flange of the floor beam.
concrete slab and the floor beams form a composite floor system. For the design loads after dening of concrete, the transformed section is used to check the stresses.
construction sequence is as follows:
The structural steel floor (floor beam, metal deck, and shear connectors) is fabricated in the shop, brought to the floor location, and placed in position. In some cases, the beams and deck are preassembled and placed as a module.
The metal deck is used as the formwork, and concrete is poured on the metal deck. Until concrete hardens, the load is carried by the metal deck and the steel floor beam.
During concreting, no shoring is provided.]*
ign Approach e floor design considers the dead, live, construction, extreme environmental, and other applicable s identified in Section 3H.3.3. The design floor loading includes the equipment attached to the
- r. The end condition for the steel beams is simply supported, or continuous. The seismic load is ined using the applicable floor acceleration response spectrum (7 percent damping for the SSE s).
load combinations applicable to the design of these floors are shown in Tables 3.8.4-1 3.8.4-2. The design of the floor system is performed in two parts:
Design of structural steel beams
- The structural steel floor beams are evaluated to withstand the weight of wet concrete during the placement of concrete. The composite section is designed for the design loads during normal and extreme environment conditions. Shear connectors are also designed.
Design of concrete slab
- The concrete slab and the steel reinforcement of the composite section are evaluated for normal and extreme environmental conditions. The slab concrete and the reinforcement is designed to meet the requirements of American Concrete Institute standard ACI 349-01 "Code Requirements for Nuclear Safety-Related Structures."
- The slab design considers the in-plane and out-of-plane seismic forces. The global in-plane and out-of-plane forces are obtained from the response spectrum analysis of the 3D finite element model of the auxiliary and shield buildings. The out-of plane seismic forces due to floor self-excitation are determined by hand calculations using the applicable vertical seismic response spectrum and slab frequency.]*
Staff approval is required prior to implementing a change in this information.
3H-13 Revision 1
e layout of this segment of the roof is shown in Figure 3H.5-7 as Region "B." The concrete slab is nches thick, plus 4.5-inch deep metal deck ribs. It is composite with 5 feet deep plate girders, ced 14-2 center to center, by using shear connectors. The girder flanges are 20 x 2 and the is 56 x 7/16. The girders span approximately 64 feet in the north-south direction and are igned as simply supported. The concrete slab between the girders behaves as a one-way slab is designed to span between the girders.
roof girders are designed for dead and live loads, including construction loads (with wet crete) with simple support end conditions. A one-third increase in allowable stress is permitted for construction load combination.
girders are also evaluated as part of the composite beam after drying of concrete. The composite structure is designed to withstand dead and live load / snow load, as well as the wind, tornado seismic loads.
pical connection of the roof slab to the shield building is shown in Figure 3H.5-7. The figure ws the arrangement of reinforcement at the connection in the fuel building roof, the shield building ndrical wall, and the walls of the auxiliary building just below the roof. The design summary is wn in Table 3H.5-10.]*
5.2.2 Floor at Elevation 135-3, Area 1 (Between Column Lines M and P) e design of a typical composite floor is shown in Figure 3H.5-6. The design summary is shown in le 3H.5-11. The concrete slab is 9 inches thick, plus 3-inch deep metal deck ribs. The floor beams typically W14x26.
The floor beams are designed for construction load (with wet concrete) with simple support end conditions. The design loads include the dead load and a construction live load of 100 pounds per square foot (psf) distributed load plus 5000 pounds concentrated load near the point of maximum shear and moment. A one-third increase in allowable stress is permitted.
The floor beams are also designed as part of the composite beam after drying of the concrete. Because of continuity of rebars into the wall and the connection of the bottom flange to the support embedment, the end support condition is considered as fixed.]*
5.3 Reinforced Concrete Slabs nforced concrete floors in auxiliary building are 24 inch or 36 inch thick. These floors are structed with 16 or 28 of reinforced concrete placed on the top of 8 inch thick precast concrete els. The 8 thick precast concrete panels are installed at the bottom to serve as the formwork and stand the load of wet concrete slab. The main reinforcement is provided in the precast panels ch are connected to the concrete placed above it by shear reinforcement. The precast panels and cast-in-place concrete act together as a composite reinforced concrete slab. Examples of such rs are the Operations Work Area (Tagging Room) ceiling slab at elevation 135 ft 3 inches in a 2, and the Area 5/6 elevation 100-0 slab between column lines 1 & 2.]*
Staff approval is required prior to implementing a change in this information.
3H-14 Revision 1
cal cross section and reinforcement. The design summary is shown in Table 3H.5-12. Design ensions of the Operations Work Area (Tagging Room) Ceiling are as follows:
om Size: 16-0 x 11-10 undary Conditions: Fixed at Walls J and K ar Span: 16-0 b Thickness: Total = 24 inches Precast Panel = 8 inches Cast-in-Place = 16 inches two precast concrete panels, each 5-11 wide and spanning over 16-0 clear span, are installed erve as the formwork.]*
5.4 Concrete Finned Floors e ceilings of the main control room and the instrumentation and control rooms in the auxiliary ding are designed as finned-floor modules. A typical floor design is shown in Figure 3H.5-9. A ed floor consists of a 24-inch-thick concrete slab poured over a stiffened steel plate ceiling. The welded to stiffen the steel plate, are half inch by 9 inch rectangular sections perpendicular to the
- e. Shear studs are welded on the other side of the steel plate, and the steel and concrete act as a posite section. The fins are exposed to the environment of the room and enhance the heat-orbing capacity of the ceiling. Several shop-fabricated steel panels, cut to room width and placed by side perpendicular to the room length, are used to construct the stiffened plate ceiling in a ularized fashion. The stiffened plate with fins is designed to withstand construction loads prior to crete hardening.
main control room ceiling fin floor is designed for the dead, live, and the seismic loads. The ign summary is shown in Table 3H.5-13.
finned floor structure is evaluated for the load combinations listed in Tables 3.8.4-1 and 3.8.4-2.]*
ign Methodology e finned floors are designed as reinforced concrete slabs in accordance with ACI Standard 349.
positive bending, the steel plate is in tension. The steel plate with fin stiffeners serves the function ottom rebars. For negative bending, the potential for buckling due to compression in this element hecked by using the criteria of American National Standards Institute/American Institute of Steel struction standards ANSI/AISC N690-94. Twisting, and therefore lateral buckling of the stiffener, strained by the concrete.
finned floors resist vertical and in-plane forces for both normal and extreme loading conditions.
positive bending, the concrete above the neutral axis carries compressive stresses and the ened steel plate resists tension. Negative bending compression is resisted by the stiffened plate tension by top rebars in the concrete. The neutral axis for negative bending is located in the ened plate section, and the concrete in tension is assumed inactive. Horizontal in-plane forces resisted by the stiffened plate and longitudinal rebars.
imum top reinforcement is provided in the slab in each direction for shrinkage and temperature k control. In addition, top reinforcement located parallel to the stiffeners is used as tension Staff approval is required prior to implementing a change in this information.
3H-15 Revision 1
posite section properties, based on an all steel-transformed section, as detailed in tion Q1.11 of ANSI/AISC N690-94, are used to design the following:
Weld strength between stiffener and the steel plate Spacing of the shear studs for the composite action stiffened plate alone is designed to resist all construction loads prior to the concrete hardening.
plate is designed against the criteria for bending and shear, specified in ANSI/AISC N690-94, tions Q1.5.1.4 and Q1.5.1.2. In addition, the weld between the stiffener and the steel plate is igned to satisfy the code requirements.]*
5.5 Structural Modules uctural modules are used for some of the structural elements on the south side of the auxiliary ding. These structural modules are structural elements built up with welded steel structural pes and plates. The modules consist of steel faceplates connected by steel trusses as shown in re 3.8.3-2. The primary purpose of the trusses is to stiffen and hold together the faceplates ng handling, erection, and concrete placement. The thickness of the steel faceplates is 0.5 inch ept in a few local areas. The nominal spacing of the trusses is 30 inches. Shear studs are welded e inside faces of the steel faceplates. Faceplates are welded to adjacent faceplates with full etration welds so that the weld is at least as strong as the plate. The structural wall modules are hored to the concrete base by reinforcing steel dowels or other types of connections embedded in reinforced concrete below. After erection, concrete is placed between the faceplates.
se modules include the spent fuel pool, fuel transfer canal, and cask loading and cask washdown The structural modules are similar to the structural modules for the containment internal ctures (see description in subsection 3.8.3 and Figures 3.8.3-8, 3.8.3-14, 3.8.3-15 and 3.8.3-17).
re 3.8.4-5 shows the location of the structural modules in the auxiliary building. The structural ules extend from elevation 66-6 to elevation 135-3.
loads and load combinations applicable to the structural modules in the auxiliary building are the e as for the containment internal structures]* (Subsection 3.8.3.5.3) [except that there are no S nor pressure loads due to pipe breaks.
design methodology of these modules in the auxiliary building is similar to the design of the ctural modules in the containment internal structures]* described in Subsection 3.8.3.5.3.
5.5.1 West Wall of Spent Fuel Pool ure 3H.5-10 shows an elevation of the west wall of the spent fuel pool (column line L-2), and ment numbers in the finite element model. The wall is a 4 feet thick concrete filled structural wall ule.
ite element analysis is performed for seismic, thermal, and hydrostatic loads with the following umptions:
The seismic in-plane and out-of-plane forces are obtained from the response spectrum analysis of the 3D finite element model of the auxiliary and shield buildings.
The thermal loads are applied as linearly varying temperatures between the inner and outer faces of the walls and floors.
Staff approval is required prior to implementing a change in this information.
3H-16 Revision 1
The seismic sloshing is modeled in the spent fuel pool.
concrete filled structural wall modules are designed as reinforced concrete structures in ordance with the requirements of ACI-349. The face plates are treated as reinforcing steel.
hods of analysis are based on accepted principles of structural mechanics and are consistent the geometry and boundary conditions of the structures. Both computer codes and hand ulations are used.
le 3H.5-8 shows the required plate thickness for certain critical locations. The steel plates are inch thick.]*
5.6 Shield Building Roof and Connections e shield building roof is a reinforced concrete shell (supporting the passive containment cooling em tank and air diffuser), which is supported on a structural steel module. The structural figuration is shown on sheets 7, 8, and 9 of Figure 3.7.2-12. Air intakes are located at the top of cylindrical portion of the shield building. The conical roof supports the passive containment ling system tank. The conical roof is constructed as a structural steel module and lifted into place ng construction. Steel beams provide permanent structural support for steel liner and concrete.
concrete is cast in place. Connection between concrete and steel liner are made using shear s.
design of the shield building is shown in Figure 3H.5-11 (Sheets 1-6). These figures show the cal details of the Tension Ring, the Air Inlet Structure, and the Exterior Wall of the Passive tainment Cooling System Tank. Figure 3H.5-16, Sheets 1 and 2, also shows the typical ensions of the surface plates and the SC to RC connections on the shield building cylindrical ment.
etailed ANSYS model was used to represent these components of the enhanced design.
lyses were performed to determine the response of the structures for the dead weight, rostatic load due to PCS water, snow load, wind load, tornado load, seismic load (including mic-induced pressure on PCS wall), and thermal loads. The design was evaluated to comply with requirements of ANSI/AISC N690-94 and of ACI 349-01.
design summaries of the components are included in Table 3H.5-9.
steel frame for the shield building roof and the concrete placed directly thereon is designed to C N690.
In the radial direction, the steel beams, the steel surface plate, and the concrete are evaluated as a composite section using the axial and bending member forces in the steel and concrete section from the finite element analyses. The steel stresses and the end connection are calculated assuming the steel alone resists all loads applied before the concrete has reached 75 percent of its required strength and the effective composite section resists all loads applied after that time.
The concrete is evaluated using all member forces in the concrete and surface steel plate from the finite element analyses (in-plane and out-of-plane forces and moments). The circumferential channels are provided for construction only and are not modeled in the finite Staff approval is required prior to implementing a change in this information.
3H-17 Revision 1
itional information is provided in Table 3H.5-15.]*
5.6.1 Air Inlets and Tension Ring e configuration and plate size of the air inlets enhance their structural performance. The air inlets cture (as shown on Figure 3H.5-14) is located at the top of the cylindrical wall portion of the shield ding, beginning at approximately elevation 251 and rising to approximately elevation 266. The air s serve as the intake for air as part of the PCS.
ve the air inlets, at approximately elevation 266, is the connection designated as the tension ring connects and supports the conical roof. The tension ring also contains 32 radial beam seat nections where the W36 x 393 radial beams for the conical roof are connected.
air inlets region is 4.5-feet thick with steel plates on each face as the primary reinforcement, ch are connected using tie bars. Near the bottom of the air inlet structure, the thickness transitions feet thick to connect with the shield building cylinder. The air inlet openings are formed using at a downward inclination of 38 degrees from the vertical. The pipe spacing is approximately degrees circumferentially with shear studs welded to the outside surface of the pipes. The tie are located with three bars between adjacent air inlets at each elevation at maximum design cing of 8.5 inches vertically. At approximately the same elevations as the tie bars, two 3/4-inch by ch (minimum) shear studs are located between the tie bars except at elevations where there is rference with the air inlet pipes. Tie bars and studs may be omitted in local areas due to design ures and other obstructions.
tension ring is designed as a structural steel box structure with concrete infill and shear studs.
the connection of the RC conical roof to the tension ring is designed to be a mechanical nection. The air inlets and tension ring design methodology is supported by linear analysis and chmarked nonlinear analysis. The tension ring is designed to ANSI/AISC N690 and is a concrete-d box girder, with two continuous 1.5-inch-thick steel plates top and bottom, which connect the r liner plate to the outer liner plate, as shown in Figure 3H.5-15.]*
5.6.2 Compression Ring and Interior Wall of Passive Containment Cooling Water Storage Tank e other areas of the shield building are designed to existing industry code requirements, and ude the conical roof, the passive containment cooling water storage tank, the compression ring, knuckle region, and their related attachments. These areas are designed as RC structures in ordance with ACI 349. The steel frame for the roof is designed for the applicable building code SI/AISC N690. The concrete roof is designed to ACI 349 requirements without credit for the steel e on the bottom of the concrete. The configuration and reinforcement of the compression ring and connection to the interior wall of the passive containment cooling water storage tank is shown in re 3H.5-11.
itional information is provided in Table 3H.5-15.]*
5.6.3 Knuckle Region and Exterior Wall of Passive Containment Cooling System Tank e exterior wall of the passive containment cooling system tank is two feet thick. The wall starts at tank floor elevation of 293 9. There is a stainless steel liner on the inside surface of the tank. The Staff approval is required prior to implementing a change in this information.
3H-18 Revision 1
rior wall are the hydrostatic pressure of the water, the in-plane and out-of-plane seismic onse, and the temperature gradient across the wall. The reinforcement is shown in re 3H.5-11. The reinforcement required and the reinforcement provided is summarized in le 3H.5-9.
itional information is provided in Table 3H.5-15.]*
5.7 Shield Building Cylinder (SC) 5.7.1 Shield Building Cylindrical Wall e shield building surrounds the containment vessel and shares a common basemat with the tainment vessel and the auxiliary building. The cylindrical shield wall has an outside radius of feet and a thickness of 36 inches. The cylindrical wall section that is below the auxiliary building line is a reinforced concrete structure. The section that is not protected by the auxiliary building is eel concrete composite structure, where two 0.75-inch plates act compositely with 34.5 inches of crete via tie bars and shear studs. The steel plate modules are connected to the reinforced crete basemat and walls by mechanical connectors as described below.
pical configuration of the SC wall is shown in Figure 3H.5-13. The overall thickness of 36 inches e same as the RC wall below. The concrete for the SC portion is standard concrete with a pressive strength of 6000 psi. The SC portion is constructed with steel surface plates, which act oncrete reinforcement. The nominal thickness of the steel faceplates is 0.75 inches. In each ule, tie bars are welded to the steel faceplates to develop composite behavior of the steel plates and concrete. The shear studs are welded to the inside surface of the steel plate to ide composite action. The tie bars are at closer spacing in the higher stress regions. The forcement detailing incorporates ACI 349 requirements.
panels of the SC wall are welded together with a complete joint penetration weld.
wall is designed for the applicable loads described in subsection 3H.3.3. A finite element lysis is performed to determine the design forces.
le 3H.5-14 shows the design summary for the enhanced shield SC cylindrical wall. The three ets represent locations in the shield building cylinder that have some of the largest demands due echanical loads. The element on the west side at grade near the RC/SC connection has large ion forces due to overturning of the cylinder under seismic demand. This area is one of the most ssed elements in tension. The element near the fuel handling building roof at elevation 180 is an ment with high out-of-plane shear due to the interaction between the fuel handling building and the nder during an earthquake. This element is located close to the fuel building roof. The element ve wall 7.3 at elevation 175 has the largest demand for out-of-plane shear in the general part of cylindrical wall away from the SC/RC connection and the interface with the auxiliary building roof.
itional discussion and information are provided in Section 4 and Figures 5 and 6 of
-GW-GLR-602 (Reference 1).]*
Staff approval is required prior to implementing a change in this information.
3H-19 Revision 1
e steel plate modules are anchored to the RC basemat and walls of the shield building by hanical rebar connections. The connectors provide for the direct transfer of forces from the RC forcing steel to the SC liner plates.
he horizontal connection at the interface with the RC structure that occurs on the bottom of the est SC wall module, each vertical reinforcing bar in the RC basemat wall is connected to a hanical coupler. A similar vertical connection occurs on the vertical edges of SC wall modules interface with the RC portion of the shield building wall. In the vertical connection, each hoop forcing bar in the RC wall is connected to a mechanical coupler and forces are transferred directly the hoop bars to the SC liner plate. The mechanical connections are designed to the stress s of ANSI/AISC N690 for loads in the reinforcing bars equivalent to 125 percent of the specified d strength of the weaker of the steel plate or reinforcing bar and are proven components used in ting structures. This design basis exceeds the maximum demand that occurs on the west side of shield building at grade and is summarized in Sheet 3 of Table 3H.5-14. This connection roves the overall ductility of the RC/SC connection.
itional discussion and information are provided in Section 4 and Figures 1, 2, 3, and 4 of
-GW-GLR-602 (Reference 1).]*
5.8 References
[APP-GW-GLR-602, Revision 1 (Proprietary) and APP-GW-GLR-603, Revision 1 (Non-Proprietary), AP1000 Shield Building Design Details for Select Wall and RC/SC Connections, Westinghouse Electric Company LLC.]*
Staff approval is required prior to implementing a change in this information.
3H-20 Revision 1
Structure (See detail in Subsection 3H.3.3.) Load Temperature (°F) Remark
[(Outside) (Inside)
S Tank Walls Normal Thermal, To -40 +40 -
+115 +70]*
[(Outside) (Inside) fs and Exterior Normal Thermal, To -40 +70 -
ls Above Grade +115 +70 Temperatures Accident Thermal, Ta -40 +132 MSIV room
-40 +212]* Fuel handling area
[(Outside) (Inside) fs and Exterior Normal Thermal, To -21.6 +47 24 thickness ls Above Grade -22.8 +48.4 27 thickness crete Temperatures -25.4 +51.5 36 thickness
+3.2 +46.6 15 insulated roof
+109.1 +79.2 24 thickness
+108.0 +80.7 27 thickness
+107.5 +81.3 36 thickness
+98.6 +81.3 15 insulated roof Accident Thermal, Ta -40 +132 MSIV room
-40 +212 Fuel handling area
+63 +212]* Insulated roof
[(Side 1) (Side 2) rior Walls/Slabs Normal Thermal, To N/R N/R -
crete Temperatures Accident Thermal, Ta +70 +132 MSIV room
+70 +212]* Fuel handling area erior Walls Below Grade Normal Thermal, To N/R N/R -
Accident Thermal, Ta N/R N/R -
emat Normal Thermal, To N/R N/R -
Accident Thermal, Ta N/R N/R -
[(Outside) (Inside) eld Building Normal Thermal, To -40 +70 -
tween Upper Annulus and +115 +70 iliary Building)
Accident Thermal, Ta -40 +132 MSIV room wall N/R N/R]* Rest of wall s:
N/R means loads due to a thermal gradient are not required to be considered.
Based on ACI 349-01 (Appendix A), the base temperature for the construction is assumed to be 70°F.
Staff approval is required prior to implementing a change in this information.
3H-21 Revision 1
(Units: kips, ft) ad Combination MX MY MXY TX TY TXY ation 180-0 to 135-3 L + H + Ta 177.8 3.1 115.5 8.8 D + 1.3 L + 1.3 H 106.4 5.6 117.0 23.9 To]*
ation 135-3 to 100-0 L + H + Ta 50.8 0.3 89.8 104.8 L + H + Ta 82.9 7.6 172.9 24.8 L + H + Ta]* 60.0 3.6 165.7 106.0 ation 100-0 to 82-6 D + 1.3 L + 1.3 H 48.1 8.4 106.1 17.3 To L+ Es]* 1.8 5.4 15.6 58.6 ation 82-6 to 66-6 L - Es 93.8 26.5 170.7 31.5
+ Es 32.7 27.2 182.1 42.4
+ Es]* 15.5 27.2 18.6 42.4 long the horizontal direction, and Y is in the vertical direction.
Staff approval is required prior to implementing a change in this information.
3H-22 Revision 1
(See Figure 3H.5-2 for Locations of Wall Sections.)
Wall Segment Required(2) [Provided (Minimum)]*
(See detail in ubsection 3H.5.1.1.) Location Vertical Horizontal Shear Vertical Horizontal Shear l Section 1, 6 ation 180-0 to 135-3 NR None Outside Face 3.48 2.65 [3.91 3.12 Inside Face 1.94 1.52 3.12 3.12]*
l Section 2, 3, 7 ation 135-3 to 100-0 NR None Outside Face 1.88 3.04 [3.12 3.12 Inside Face 1.77 2.23 3.12 3.12]*
l Section 4, 8 ation 100-0 to 82-6 0.003 [0.44]*
Outside Face 1.42 0.70 [3.12 1.56 Inside Face 1.01 0.70 3.12 1.27]*
l Section 5, 9 ation 82-6 to 66-6 0.27 [1.00]*
Outside Face 2.29 0.87 [4.39 1.27 Inside Face 1.87 0.87 3.12 1.27]*
NR = not required.
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-23 Revision 1
(Units: kips, ft)
Load Combination MX MY MXY TX TY TXY m Roof to Elevation 155-6 D + 1.3 L + 1.2 To 135.3 10.9 117.3 210.2 D + 1.3 L + 1.2 To]* 75.5 4.1 229.8 94.3 ation 155-6 to 135-3 D - Es 14.1 1.3 160.8 228.7 L - Es]* 28.0 1.0 29.8 231.7 ation 135-3 to 117-6 D - Es 3.3 1.3 142.2 140.9 L - Es]* 10.0 1.0 41.7 175.0 ation 117-6 to 100-0 D - Es 4.7 2.8 143.9 184.9 L + Es]* 6.4 1.5 172.8 107.9 ation 100-0 to 82-6 D - Es 15.4 2.6 90.4 169.8 L - Es]* 8.7 2.6 46.6 175.6 ation 82-6 to 66-6 D - Es 23.5 1.3 80.9 49.3 L - Es]* 0.8 1.3 1.7 74.1 long the horizontal direction, and Y is in the vertical direction.
Staff approval is required prior to implementing a change in this information.
3H-24 Revision 1
(See Figure 3H.5-2 for Locations of Wall Sections.)
Wall Segment Reinforcement on Each Face (in2/ft)
(See detail in Subsection 3H.5.1.2.) Location Wall Section Required(1) [Provided (Min.)]*
m Roof to Elevation 155-6 Horizontal 1 3.96 [4.12 Vertical 7 3.60 3.72 vation 155-6 to 135-3 Horizontal 2 2.80 3.12 Vertical 8 3.59 3.72 vation 135-3 to 117-6 Horizontal 3 2.03 2.54 Vertical 9 2.63 3.12 vation 117-6 to 100-0 Horizontal 4 2.29 2.54 Vertical 10 2.98 3.12 vation 100-0 to 82-6 Horizontal 5 1.69 2.54 Vertical 11 2.08 3.12 vation 82-6 to 66-6 Horizontal 6 0.85 1.27 Vertical 12 0.98 1.56 2/ft2) ar Reinforcement (in m Roof to Elevation 155-6 Standard hook or 7 0.38 0.44]*
T headed bar Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-25 Revision 1
(Units: kips, ft) oad Combination MX MY MXY TX TY TXY ation 154-2 to 135-3 D + Es+ Pa + Yj 6.0 3.5 115.4 170.2 D + Es+ Pa + Yj]* 14.3 3.5 46.0 170.2 ation 135-3 to 117-6 D + Es+ Pa + Yj 145.3 12.2 26.0 38.2 D + Es+ Pa + Yj]* 24.5 7.1 15.5 114.9 along the horizontal direction, and Y is in the vertical direction.
Staff approval is required prior to implementing a change in this information.
3H-26 Revision 1
(See Figure 3H.5-2, Sheet 3, for Locations of Wall Sections.)
Wall Segment Reinforcement on Each Face (in2/ft2)
(See detail in Wall Subsection 3H.5.1.3.) Location Section Required(1) [Provided (Min.)]*
ation 154-2 to 135-3 Horizontal 1 2.08 [2.27 Vertical 3 2.59 3.12 ation 135-3 to 117-6 Horizontal 2 1.36 4.39 Vertical 4 2.02 5.66]*
2/ft2) ar Reinforcement (in ation 154-2 to 135-3 Standard hook or 5 0.01 [0.11 T headed bar ation 135-3 to 117-6 Standard hook or 6 0.33 2.00]*
T headed bar Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-27 Revision 1
Comparisons to Acceptance Criteria - Element No. 20477 Load/ Sxx Syy Sxy Mxx Myy Nx Ny bination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d (D) -16.15 -22.92 -28.34 -1.34 -1.06 -0.32 -0.32 (L) 1.46 0.32 -1.57 -0.06 -0.21 0.04 0.03 o (F) 37.52 12.36 -4.32 -100.50 -14.49 62.14 -9.95 mic (Es) 46.21 56.51 183.20 81.72 28.70 103.00 14.79 mal (To) -561.80 -267.70 -51.15 -426.90 -145.50 90.32 -23.66 mal (Ta) -955.80 -444.60 -139.70 -1401.0 -450.00 227.50 -83.16 a) 32.40 -14.25 -48.39 -142.68 -22.12 86.61 -14.33 [1.4D+1.7L+1.4F a) 84.05 51.21 147.24 -60.38 7.15 189.71 0.56 D+L+F+Es b) 84.05 51.21 -219.16 -223.82 -50.25 -16.29 -29.02 D+L+F+E's e) -267.08 -116.11 115.28 -327.19 -83.79 246.16 -14.22 D+L+F+Es+To f) -267.08 -116.11 -251.12 -490.63 -141.19 40.16 -43.80 D+L+F+E's+To m) 84.20 53.18 151.64 -60.18 7.46 189.71 0.57 0.9D+F+Es n) 84.20 53.18 -214.76 -223.62 -49.94 -16.29 -29.01 0.9D+F+E's o) -266.92 -114.13 119.68 -326.99 -83.47 246.16 -14.22 0.9D+F+Es+To p) -266.92 -114.13 -246.72 -490.43 -140.87 40.16 -43.80 0.9D+F+E's+To a) -574.55 -288.12 -121.54 -977.52 -297.00 204.04 -62.22 D+L+F+Ta b) -825.30 -421.18 -153.29 -53.19 -5.28 63.89 -15.73 D+L+F+Ta a) -397.01 -211.45 -74.69 -427.19 -125.72 132.70 -28.49 1.05D+1.3L+1.05F+1.2To]*
s:
irection is horizontal; y - direction is vertical.
Figure 3H.5-10 for element location.
thickness required for load combinations excluding thermal: 0.42 inches (Maximum) e thickness provided: 0.50 -0.01 +0.10 inches]*
mum principal stress for load combination 5 including thermal: 46.33 ksi stress: 65.0 ksi (Minimum)]*
mum stress intensity range for load combination 5 including thermal: 46.3 ksi able stress intensity: 130.0 ksi (Minimum)
Staff approval is required prior to implementing a change in this information.
3H-28 Revision 1
Comparisons to Acceptance Criteria - Element No. 10529 Load/ Sxx Syy Sxy Mxx Myy Nx Ny bination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments (D) -24.40 -96.30 -20.71 -1.16 -2.27 -0.28 -0.34 L) -0.44 -2.48 -0.55 -0.01 -0.24 0.01 0.08 o (F) 9.86 -5.49 6.22 8.37 -73.49 16.94 16.02 mic (Es) 110.80 335.20 95.73 19.03 93.81 22.15 29.34 mal (To) -215.70 -479.30 -150.10 -99.69 -357.90 16.39 19.34 mal (Ta) -389.40 -883.60 -273.20 -364.10 -982.20 40.42 17.26 a) -21.10 -146.72 -21.23 10.09 -106.48 23.34 22.09 [1.4D+1.7L+1.4F a) 99.77 228.74 83.17 29.58 -11.59 45.60 51.51 D+L+F+Es b) 99.77 228.74 -108.29 -8.48 -199.21 1.30 -7.17 D+L+F+E's e) -35.05 -70.83 -10.64 -32.72 -235.28 55.84 63.60 D+L+F+Es+To
) -35.05 -70.83 -202.10 -70.78 -422.90 11.54 4.92 D+L+F+E's+To m) 102.64 240.85 85.80 29.71 -11.12 45.61 51.47 0.9D+F+Es n) 102.64 240.85 -105.66 -8.35 -198.74 1.31 -7.21 0.9D+F+E's o) -32.17 -58.72 -8.02 -32.60 -234.81 55.86 63.55 0.9D+F+Es+To p) -32.17 -58.72 -199.48 -70.66 -422.43 11.56 4.87 0.9D+F+E's+To a) -258.35 -656.52 -185.79 -220.36 -689.88 41.93 26.55 D+L+F+Ta b) -362.67 -963.64 -260.17 7.94 -144.07 12.21 12.80 D+L+F+Ta a) -177.61 -469.58 -128.51 -67.20 -348.29 29.80 31.07 1.05D+1.3L+1.05F+1.2To]*
s:
irection is horizontal; y - direction is vertical.
Figure 3H.5-10 for element location.
thickness required for load combinations excluding thermal: 0.47 inches (maximum) e thickness provided: 0.50 -0.01 +0.10 inches]*
mum principal stress for load combination 5 including thermal: 40.3 ksi stress: 65.0 ksi (Minimum]*
mum stress intensity range for load combination 5 including thermal: 50.8 ksi able stress intensity: 130.0 ksi (Minimum)
Staff approval is required prior to implementing a change in this information.
3H-29 Revision 1
Comparisons to Acceptance Criteria - Element No. 10544 Load/ Sxx Syy Sxy Mxx Myy Nx Ny bination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d (D) -20.03 -75.69 -42.72 3.53 -2.18 -0.01 -1.93 (L) -0.64 -1.98 -1.22 0.36 -0.06 0.02 -0.07 o (F) -4.13 -2.97 -4.10 39.78 3.54 0.99 -4.80 mic (Es) 67.42 185.70 113.20 48.28 7.62 5.78 5.32 mal (To) -121.60 -387.30 -239.80 75.83 -107.40 39.64 49.91 mal (Ta) -215.20 -670.10 -416.60 184.20 -269.30 115.50 136.20 a) -34.91 -113.49 -67.62 61.25 1.81 1.40 -9.54 [1.4D+1.7L+1.4F a) 40.97 103.87 63.52 107.86 10.34 7.18 -3.41 D+L+F+Es b) 40.97 103.87 -162.88 11.30 -4.90 -4.39 -14.04 D+L+F+E's e) -35.03 -138.19 -86.36 155.26 -56.79 31.95 27.79 D+L+F+Es+To f) -35.03 -138.19 -312.76 58.70 -72.02 20.39 17.15 D+L+F+E's+To m) 43.61 113.42 69.01 107.15 10.61 7.16 -3.14 0.9D+F+Es n) 43.61 113.42 -157.39 10.59 -4.62 -4.41 -13.78 0.9D+F+E's o) -32.39 -128.64 -80.87 154.54 -56.51 31.93 28.05 0.9D+F+Es+To p) -32.39 -128.64 -307.27 57.98 -71.75 20.37 17.41 0.9D+F+E's+To a) -159.30 -499.45 -308.41 158.79 -167.01 73.19 78.32 D+L+F+Ta b) -267.05 -805.64 -503.54 51.38 -38.58 1.37 -9.65 D+L+F+Ta a) -117.40 -375.64 -230.60 102.82 -79.20 30.78 30.27 1.05D+1.3L+1.05F+1.2To]*
s:
irection is horizontal; y - direction is vertical.
Figure 3H.5-10 for element location.
thickness required for load combinations excluding thermal: 0.31 inches (Maximum) e thickness provided: 0.50 -0.01 +0.10 inches]*
mum principal stress for load combination 5 including thermal: 46.95 ksi stress: 65.0 ksi (Minimum)]*
mum stress intensity range for load combination 5 including thermal: 84.9 ksi able stress intensity: 130.0 ksi (Minimum)
Staff approval is required prior to implementing a change in this information.
3H-30 Revision 1
Comparisons to Acceptance Criteria - Element No. 10524 Load/ Sxx Syy Sxy Mxx Myy Nx Ny bination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d (D) -35.61 -104.80 0.68 -4.70 7.72 -0.55 -2.22 (L) -0.45 -2.21 -0.72 -0.25 -0.49 0.00 0.10 ro (F) 11.85 -1.35 4.92 28.52 16.50 3.71 3.79 mic (Es) 76.80 225.60 79.29 53.31 177.00 6.83 55.70 rmal (To) -369.10 -433.40 179.90 -215.40 -109.40 -7.32 -59.63 rmal (Ta) -696.60 -730.00 329.40 -555.10 -487.60 -13.58 -95.78 1a) -34.04 -152.37 6.62 32.92 33.09 4.43 2.37 [1.4D+1.7L+1.4F 3a) 57.33 116.69 86.14 88.29 207.34 11.48 58.89 D+L+F+Es 3b) 57.33 116.69 -72.44 -18.33 -146.66 -2.18 -52.51 D+L+F+E's e) -173.36 -154.18 198.57 -46.34 138.96 6.90 21.62 D+L+F+Es+To 3f) -173.36 -154.18 39.99 -152.96 -215.04 -6.76 -89.78 D+L+F+E's+To 3m) 61.34 129.38 86.78 89.00 207.05 11.53 59.02 0.9D+F+Es 3n) 61.34 129.38 -71.80 -17.62 -146.95 -2.13 -52.38 0.9D+F+E's o) -169.35 -141.49 199.22 -45.62 138.68 6.96 21.75 0.9D+F+Es+To 3p) -169.35 -141.49 40.64 -152.24 -215.32 -6.71 -89.65 0.9D+F+E's+To 5a) -459.59 -564.62 210.75 -323.37 -281.01 -5.32 -58.19 D+L+F+Ta 5b) -741.71 -755.24 398.88 19.86 124.99 -105.77 -114.64 D+L+F+Ta 7a) -302.36 -439.4 139.9 136.9 57.2 -2.2 -42.9 1.05D+1.3L+1.05F+1.2To]*
s:
irection is horizontal; y - direction is vertical.
Figure 3H.5-10 for element location.
thickness required for load combinations excluding thermal: 0.32 inches (Maximum) e thickness provided: 0.50 -0.01 +0.10 inches]*
mum principal stress for load combination 5 including thermal: 42.1 ksi stress: 65.0 ksi (Minimum)]*
mum stress intensity range for load combination 5 including thermal: 72.5 ksi able stress intensity: 130.0 ksi (Minimum)
Staff approval is required prior to implementing a change in this information.
3H-31 Revision 1
Comparisons to Acceptance Criteria - Element No. 20462 Load/ Sxx Syy Sxy Mxx Myy Nx Ny bination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d (D) -7.31 -29.13 -1.51 -1.45 -3.75 -0.06 0.35 (L) -0.11 -0.55 0.21 -0.14 -0.60 0.00 0.05 o (F) 5.04 -0.04 -1.61 -16.58 64.59 -1.48 -20.87 mic (Es) 25.64 33.82 32.90 10.45 114.90 2.48 12.55 mal (To) -286.10 -78.70 66.37 -208.70 -130.00 0.86 -1.51 mal (Ta) -616.80 -121.80 116.60 -650.20 -502.40 6.16 3.93 a) -3.36 -41.77 -4.01 -25.47 84.16 -2.15 -28.64 [1.4D+1.7L+1.4F a) 25.28 4.09 29.35 -14.35 200.98 0.35 -16.27 D+L+F+Es b) 25.28 4.09 -36.45 -35.25 -28.82 -4.61 -41.37 D+L+F+E's e) -153.54 -45.10 70.83 -144.78 119.73 0.89 -17.21 D+L+F+Es+To f) -153.54 -45.10 5.03 -165.68 -110.07 -4.07 -42.31 D+L+F+E's+To m) 26.11 7.55 29.29 -14.06 201.95 0.35 -16.35 0.9D+F+Es n) 26.11 7.55 -36.51 -34.96 -27.85 -4.61 -41.45 0.9D+F+E's o) -152.70 -41.63 70.77 -144.50 120.70 0.89 -17.29 0.9D+F+Es+To p) -152.70 -41.63 4.97 -165.40 -109.10 -4.07 -42.39 0.9D+F+E's+To a) -387.88 -105.84 69.97 -424.54 -253.76 2.31 -18.01 D+L+F+Ta b) -646.13 -113.41 80.41 35.38 175.18 -4.36 -31.38 D+L+F+Ta a) -217.10 -90.37 46.78 -175.63 -34.40 -0.96 -22.61 1.05D+1.3L+1.05F+1.2To]*
s:
irection is horizontal; y - direction is vertical.
Figure 3H.5-10 for element location.
thickness required for load combinations excluding thermal: 0.20 inches (Maximum) e thickness provided: 0.50 -0.01 +0.10 inches]*
mum principal stress for load combination 5 including thermal: 20.6 ksi stress: 65.0 ksi (Minimum)]*
mum stress intensity range for load combination 5 including thermal: 20.6 ksi able stress intensity: 130.0 ksi (Minimum)
Staff approval is required prior to implementing a change in this information.
3H-32 Revision 1
Comparisons to Acceptance Criteria - Element No. 21402 Load/ Sxx Syy Sxy Mxx Myy Nx Ny bination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d (D) -1.82 -17.93 4.00 0.92 0.93 -0.32 0.22 (L) -0.21 -0.98 0.41 0.19 -0.04 -0.02 -0.03 o (F) 7.14 0.29 -2.18 104.60 15.51 -16.65 3.08 mic (Es) 36.81 21.41 17.68 139.90 28.75 12.42 12.08 mal (To) -228.50 -181.90 85.52 -291.30 -212.00 11.34 6.92 mal (Ta) -379.10 -378.40 159.80 -783.80 -661.10 41.72 28.29 a) 7.08 -26.36 3.24 148.06 22.95 -23.80 4.56 [1.4D+1.7L+1.4F a) 44.77 2.90 19.03 287.45 51.36 -11.24 16.58 D+L+F+Es b) 44.77 2.90 -16.33 7.65 -6.14 -36.08 -7.58 D+L+F+E's e) -98.05 -110.78 72.48 105.39 -81.14 -4.15 20.90 D+L+F+Es+To f) -98.05 -110.78 37.12 -174.41 -138.64 -28.99 -3.26 D+L+F+E's+To m) 45.16 5.68 18.23 287.17 51.31 -11.18 16.59 0.9D+F+Es n) 45.16 5.68 -17.13 7.37 -6.19 -36.02 -7.57 0.9D+F+E's o) -97.65 -108.01 71.68 105.11 -81.19 -4.09 20.91 0.9D+F+Es+To p) -97.65 -108.01 36.32 -174.69 -138.69 -28.93 -3.25 0.9D+F+E's+To a) -231.84 -255.12 102.10 -384.16 -396.79 9.08 20.95 D+L+F+Ta b) -268.90 -468.00 168.35 -17.41 14.23 -18.83 13.88 D+L+F+Ta a) -166.1 -156.2 66.6 -107.4 -141.8 -9.3 8.6 1.05D+1.3L+1.05F+1.2To]*
s:
irection is horizontal; y - direction is vertical.
Figure 3H.5-10 for element location.
thickness required for load combinations excluding thermal: 0.28 inches (Maximum) e thickness provided: 0.50 -0.01 +0.10 inches]*
mum principal stress for load combination 5 including thermal: 25.1 ksi stress: 65.0 ksi (Minimum)]*
mum stress intensity range for load combination 5 including thermal: 31.3 ksi able stress intensity: 130.0 ksi (Minimum)
Staff approval is required prior to implementing a change in this information.
3H-33 Revision 1
Comparisons to Acceptance Criteria - Element No. 21414 Load/ Sxx Syy Sxy Mxx Myy Nx Ny bination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d (D) 0.69 -10.62 -2.57 -0.52 -0.22 -0.03 0.12 (L) 0.18 0.12 -0.45 0.00 -0.11 -0.01 0.02 o (F) 4.25 0.56 -2.73 -27.01 -31.06 -1.46 1.82 mic (Es) 26.90 13.88 36.68 26.35 21.70 2.17 4.34 mal (To) -79.35 -40.69 49.04 -129.00 -119.30 10.01 6.90 mal (Ta) -129.60 -66.37 57.50 -374.60 -374.70 26.38 24.34 a) 7.24 -13.89 -8.19 -38.54 -43.97 -2.09 2.75 [1.4D+1.7L+1.4F a) 33.73 4.16 29.84 -11.98 -22.11 0.10 7.03 D+L+F+Es b) 33.73 4.16 -43.52 -64.68 -65.51 -4.24 -1.66 D+L+F+E's e) -15.86 -21.27 60.49 -92.61 -96.67 6.36 11.34 D+L+F+Es+To f) -15.86 -21.27 -12.87 -145.31 -140.07 2.01 2.66 D+L+F+E's+To m) 33.48 5.10 30.55 -11.93 -21.98 0.11 7.00 0.9D+F+Es n) 33.48 5.10 -42.81 -64.63 -65.38 -4.23 -1.69 0.9D+F+E's o) -16.12 -20.33 61.20 -92.56 -96.54 6.37 11.31 0.9D+F+Es+To p) -16.12 -20.33 -12.16 -145.26 -139.94 2.02 2.62 0.9D+F+E's+To a) -75.87 -51.43 30.19 -261.65 -265.57 15.00 17.17 D+L+F+Ta b) -114.31 -96.07 55.47 -35.06 -36.08 2.55 -1.61 D+L+F+Ta a) -54.08 -40.93 30.63 -125.65 -122.46 5.94 7.24 1.05D+1.3L+1.05F+1.2To]*
s:
irection is horizontal; y - direction is vertical.
Figure 3H.5-10 for element location.
thickness required for load combinations excluding thermal: 0.14 inches (Maximum) e thickness provided: 0.50 -0.01 +0.10 inches]*
mum principal stress for load combination 5 including thermal: 22.1 ksi stress: 65.0 ksi (Minimum)]*
mum stress intensity range for load combination 5 including thermal: 22.1 ksi able stress intensity: 130.0 ksi (Minimum)
Staff approval is required prior to implementing a change in this information.
3H-34 Revision 1
(Tension Ring)
Tension Ring - Axial Force and Bending Verification Seismic Maximum Maximum [Design Limit(1)
Location Stresses Maximum Steel Area for Ratio Seismic fa Stresses Fy Required(2) [Steel Area Max Required/
ion Angles L/C ksi ksi ksi (in2/ft) Provided]* Provided]*
5.625° 9 14.31 14.31 50 7.74 [Liner 1 1/2" = [0.43 + 2%]*
er 84.375° 17 12.52 18 (in2/ft) (Min)]*
0° 9 12.97 er 90° 17 11.39 Tension Ring - Shear Force and Torsion Verification Seismic Maximum Maximum [Design Limit for Location Stresses Maximum Steel Area Ratio Seismic fv Stresses Fy Required(2) [Steel Area Max Required/
ion Angles L/C ksi ksi ksi (in2/ft) Provided]* Provided]*
5.625° 17 4.83 5.52 50 5.04 [Liner 1 1/2" = [0.28 + 2%]*
er 84.375° 9 5.52 18 (in2/ft) (Min.)]*
0° 18 3.20 er 90° 11 4.00 s:
[Two percent of the value may be added to the design limit as an allowance for minor variances in analysis results.]*
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-35 Revision 1
(Air Inlet)
AIS Reinforcement Summary - Horizontal Sections Steel Area (Vertical Direction - Z Local Dir.)
Required - Seismic Load [Design Limit(1)
Locations Combinations Maximum for Ratio (Figure 3H.5-11) (in2/ft) Required(2) Max Required/
tions Angles Seismic L/C Values (in2/ft) [Provided]* Provided]*
+6 0°-5.625° 16 1.65 84.375°-90° 8 1.41 8 0°-5.625° 16 2.10 2.10 [Liner 1" = 12 (in2/ft) [0.175 + 2%]*
84.375°-90° 8 1.69 (Min.)]*
9 0°-5.625° 16 2.10 84.375°-90° 8 1.68 11 0°-5.625° 16 1.61 1.61 [Liner 3/4" = 9 (in2/ft) [0.18 + 2%]*
84.375°-90° 24 1.21 (Min.)]*
s:
[Two percent of the value may be added to the design limit as an allowance for minor variances in analysis results.]*
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-36 Revision 1
(Air Inlet)
AIS Reinforcement Summary - Vertical Sections Locations (Figure 3H.5-11) Steel Area (Hoop Direction - Y Local Dir.)
Required - Seismic [Design Limit(1)
Load Combinations Maximum for Ratio (in2/ft) Required(2) Max Required/
tions Angles Seismic L/C Values (in2/ft) [Provided]* Provided]*
pper 0° 9 9.56 90° 17 8.32 ower 0° 9 8.14 90° 18 7.03 [Liner 1" = 12 (in2/ft) 10.04 [0.84 + 2%]*
pper 5.625° 9 10.04 (Min.)]*
84.375° 17 8.69 ower 5.625° 9 7.98 84.375° 19 6.82 s:
[Two percent of the value may be added to the design limit as an allowance for minor variances in analysis results.]*
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-37 Revision 1
(Air Inlet)
Out of Plane Shear Reinforcement Summary - AIS Required - Seismic Locations Load Combinations [Design Limit(1)
(Figure 3H.5-11) (in2/ft) Maximum for Ratio Seismic Required(2) [Steel Area Max Required/
les Sections L/C Values Sum (in2/ft) Provided]* Provided]*
Max of Vertical Sections 0.10
- 3 upper - 4 upper 1 0.10 25° Horizontal 0.00 Section 5+6 Max of Vertical sections 0.10 75° - 3 upper - 4 upper 1 0.10
° Horizontal 0.00 Section 5+6 Max of Vertical Sections 0.10
- 3 upper - 4 upper 9 0.34 25° Horizontal 0.24 Section 8 Max of Vertical Sections 0.10 75° - 3 upper - 4 upper 1 0.30 [3 #6 TIE BAR
° Horizontal 0.20 @2.8125° (41.36")
Section 8 0.34 (8 1/2" in vertical [0.63 + 2%]*
Max of Vertical Sections direction) =
0.093
- 3 lower - 4 lower 0.54 (in2/ft) (Min.)]*
0 0.22 25° Horizontal 0.127 Section 9 Max of Vertical Sections 0.183 75° - 3 lower - 4 lower 0 0.18
° Horizontal 0.000 Section 9 Max of Vertical Sections 0.167
- 3 lower - 4 lower 1 0.17 25° Horizontal 0.000 Section 11 Max of Vertical Sections 0.02 75° - 3 lower - 4 lower 0 0.02
° Horizontal 0.00 Section 11 s:
[Two percent of the value may be added to the design limit as an allowance for minor variances in analysis results.]*
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-38 Revision 1
(Exterior Wall Of Passive Containment Cooling System Tank)
Reinforcement on Each Face, in2/ft Location Ratio Wall (Figure 3H.5-11 Maximum Required/
egment Sheet 5 of 6) Required Provided (Minimum) Provided Vertical 1.37 1#11@1.2° [1.72 0.80 ottom Hoop 0.67 1#9@6" 2 0.33 Shear 0.07 1#6@1.2°x12" 0.48 0.15 Vertical 0.64 1#11@1.2° 1.72 0.37 d-height Hoop 1.85 1#9@6" 2 0.92 Vertical 0.52 1#11@1.2° 1.72 0.30 Top Hoop 0.79 1#9@6" 2]* 0.39 Staff approval is required prior to implementing a change in this information.
3H-39 Revision 1
(Near Shield Building Interface) erning Load Combination (Roof Girder) mbination Number 3 - Extreme Environmental Condition Downward Seismic Acceleration ding Moment = 7125 kips-ft responding Stress = 24.1 ksi wable Stress = 38.0 ksi ar Force = 447 kips responding Stress = 17.0 ksi wable Stress = 20.1 ksi erning Load Combination (Concrete Slab) allel to Girders ombination Numbers 3 - Extreme Environmental Condition einforcement (Each Face)
Required(1) = 1.74 in2/ft
[Provided = 2.54 in2/ft (Minimum)]*
pendicular to Girders ombination Numbers 3 - Extreme Environmental Condition einforcement (Each Face)
Required(1) = 1.68 in2/ft
[Provided = 3.12 in2/ft (Minimum)]*
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-40 Revision 1
erning Load Combination (Steel Beam) d Combination 3 - Extreme Environmental Condition Downward Seismic ding Moment =(-) 63.9 kips-ft responding Stress = 17.0 ksi wable Stress =33.26 ksi ar Force = 30.7 kips responding Stress = 8.7 ksi wable Stress = 20.1 ksi erning Load Combination (Concrete Slab) allel to the Beams oad Combination 3 - Extreme Environmental Condition Downward Seismic ending Moment =(-) 16.0 kips-ft/ft n-plane Shear =20.0 kips (per foot width of the slab) einforcement (Each Face)
Required(1) = 0.41 in2/ft
[Provided = 0.44 in2/ft (Min.)]*
pendicular to the Beams ombination Number Normal Condition ending Moment =(+) 6.66 kips-ft (per foot width of the slab) einforcement (Each Face)
Required(1) = 0.28 in2/ft
[Provided = 0.60 in2/ft (Min.)]*
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-41 Revision 1
(Operations Work Area (Previously Known As Tagging Room) Ceiling) ign of Precast Concrete Panels erning Load Combination Construction ign Bending Moment (Midspan) = 14.53 kip-ft/ft om Reinforcement (E/W Direction) equired(1) = 0.58 in2/ft Provided = 0.79 in2/ft (Min.)]*
Reinforcement (E/W Direction) equired(1) = (Minimum required by Code)
Provided = 0.20 in2/ft (Min.)]*
and Bottom Reinforcement (N/S Direction) equired(1) = (Minimum required by Code)
Provided = 0.20 in2/ft (Min.)]*
ign of 24-inch-Thick Slab erning Load Combination Extreme Environmental Condition (SSE) ign Bending Moment (E/W Direction) Midspan = 14.40 kips ft/ft ign In-plane Shear = 31.9 kips ft ign In-plane Tension = 21.9 kips ft om Reinforcement (E/W Direction) equired(1) = 0.53 in2/ft Provided = 0.79 in2/ft (Min.)]*
ign Bending Moment (E/W Direction) at Support = 28.81 kips-ft/ft ign In-plane Shear = 31.9 kips/ft ign In-plane Tension = 21.9 kips/ft Reinforcement (E/W Direction) equired(1) = 0.93 in2/ft Provided = 1.00 in2/ft (Min.)]*
ign Bending Moment (N/S Direction) = 8.47 kips ft/ft ign In-plane Shear = 31.9 kips/ft ign In-plane Tension = 27.2 kip/ ft and Bottom Reinforcement (N/S Direction) equired(1) = 0.59 in2/ft Provided = 0.79 in2/ft (Min.)]*
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-42 Revision 1
design of the bottom plate with fins is governed by the construction load.
the composite floor, the design forces used for the evaluation of a typical 9-inch-wide strip of the slab are ollows:
aximum bending moment=+35.0 (-24.4) kips-ft aximum shear force=22.3 kips design evaluation results are summarized below: (1)
The actual area of the tension steel is 9.0 in2 (Min.),]* which provides a design strength of 518.5 kips-ft bending moment capacity.
The design shear strength is 23.22 kips.
he shear studs are spaced a maximum of 9 inches c/c, in both directions.]* The calculated required spacing is 9.06 nches.
Thermal loads have been considered in the design of critical sections. The required reinforcement values shown do not include the load case where seismic and normal thermal loads are numerically combined as the normal thermal loads were assessed to be insignificant. When the seismic and normal thermal loads are numerically combined, the value of required reinforcement may increase; however, in all cases the required reinforcement is less than the provided reinforcement and thus the design of the critical section reinforcement is acceptable.
Staff approval is required prior to implementing a change in this information.
3H-43 Revision 1
mparison to Acceptance Criteria Elevation 180 Feet Near Fuel Handling Building Roof TX TY TXY MX MY MXY NX NY oad/Combination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d -6 -118 15 -25 -17 4 -6 -5 1 -1 0 0 0 0 0 0 mic 155 385 163 299 209 35 71 33
-7 -167 22 -35 -24 5 -8 -7 1.4 D + 1.7 L 150 266 179 274 192 38 65 28 D + L + Es 150 266 -147 -324 -226 -31 -76 -38 D + L + E's
-160 -504 -147 -324 -226 -31 -76 -38 D + L - Es
-160 -504 179 274 192 38 65 28 D + L - E's 150 278 177 277 193 38 66 28 0.9 D + Es 150 278 -149 -322 -224 -31 -76 -37 0.9 D + E's
) 211 369 229 453 294 64 105 33 0.9 D + E's +
To(W1)
) 226 357 234 463 302 64 108 33 0.9 D + E's +
To(W2) ection is horizontal; y-direction is vertical.
ent number: 12164 e thickness required for load combinations excluding thermal: 0.43 inches + 2%(1)]*
e thickness required for load combinations including thermal: 0.57 inches + 2%(1)]*
e thickness provided: 0.75 inches]*
ar reinforcement required for load combinations excluding thermal: 0.64 in2/ft2 + 2%(1)]*
ar reinforcement required for load combinations including thermal: 0.93 in2/ft2 + 2%(1)]*
r reinforcement provided: See [APP-GW-GLR-602, Section 4.]*
s:
[The Tier 2* designation for Plate thickness required requires NRC approval if this value is exceeded as a result of design changes or detail design adjustments identified during preparation of fabrication or construction drawings or instructions.]*
Load cases 35 and 37 are the two governing load combinations for element 12164 that include thermal and seismic loads combined numerically. W1 designates the winter conditions with the spent fuel pool at the normal operating temperature limit.
W2 designates the winter conditions with the spent fuel pool and fuel transfer canal at the normal operating temperature limit.
Es is SRSS (member forces are positive) of the SSE loads. E's is Es with all member forces except axial forces (TX, TY) reversed to negative.
Staff approval is required prior to implementing a change in this information.
3H-44 Revision 1
Acceptance Criteria Elevation 175 Feet Near Intersection With Wall 7.3 TX TY TXY MX MY MXY NX NY ad/Combination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d -6 -105 12 -6 5 1 0 2 0 -1 0 0 0 0 0 0 mic 34 325 176 38 25 13 2 8
-9 -149 17 -9 7 1 0 3 1.4 D + 1.7 L 28 219 188 31 30 14 2 10 D + L + Es 28 219 -164 -44 -20 -12 -3 -6 D + L + E's
-40 -431 -164 -44 -20 -12 -3 -6 D + L - Es
-40 -431 188 31 30 14 2 10 D + L - E's 28 230 187 32 29 14 2 10 0.9 D + Es 28 230 -166 -44 -20 -12 -3 -7 0.9 D + E's
) 77 227 186 36 -58 7 3 11 D + L + E's +
To(W1)
) 77 238 186 36 -58 7 3 11 0.9 D + E's +
To(W2) ection is horizontal; y-direction is vertical.
ent number: 11514 e thickness required for load combinations excluding thermal: 0.40 inches + 2%(1)]*
e thickness required for load combinations including thermal: 0.40 inches + 2%(1)]*
e thickness provided: 0.75 inches]*
ar reinforcement required for load combinations excluding thermal: 0.07 in2/ft2 + 2%(1)]*
ar reinforcement required for load combinations including thermal: 0.08 in2/ft2 + 2%(1)]*
r reinforcement provided: See [APP-GW-GLR-602, Section 4.]*
s:
[The Tier 2* designation for Plate thickness required requires NRC approval if this value is exceeded as a result of design changes or detail design adjustments identified during preparation of fabrication or construction drawings or instructions.]*
Load cases 19 and 37 are the two governing load combinations for element 11514 that include thermal and seismic loads combined numerically. W1 designates the winter conditions with the spent fuel pool at the normal operating temperature limit.
W2 designates the winter conditions with the spent fuel pool and fuel transfer canal at the normal operating temperature limit. Es is SRSS (member forces are positive) of the SSE loads. E's is Es with all member forces except axial forces (TX, TY) reversed to negative.
Staff approval is required prior to implementing a change in this information.
3H-45 Revision 1
Acceptance Criteria Elevation Grade on West Side TX TY TXY MX MY MXY NX NY ad/Combination kip/ft kip/ft kip/ft k-ft/ft k-ft/ft k-ft/ft kip/ft kip/ft Comments d 2 -127 0 2 19 0 0 -2 0 1 0 0 0 0 0 0 smic 58 477 231 2 16 17 4 7 2 -176 -1 3 26 0 0 -3 1.4 D + 1.7 L 60 352 231 4 35 18 4 5 D + L + Es 60 352 -232 1 2 -17 -4 -9 D + L + E's
-57 -603 -232 1 2 -17 -4 -9 D + L - Es
-57 -603 231 4 35 18 4 5 D + L - E's 60 364 231 4 33 18 4 5 0.9 D + Es 60 364 -232 0 1 -17 -4 -9 0.9 D + E's
)
182 364 -238 113 155 -17 -4 -31 D + L + E's +
To(W1)
) 182 380 -238 113 153 -17 -4 -31 0.9 D + E's +
To(W2) ection is horizontal; y-direction is vertical.
ent number: 23752 e thickness required for load combinations excluding thermal: 0.56 inches + 2%(1)]*
e thickness required for load combinations including thermal: 0.58 inches + 2%(1)]*
e thickness provided: 0.75 inches]*
ar reinforcement required for load combinations excluding thermal: 0.06 in2/ft2 + 2%(1)]*
ar reinforcement required for load combinations including thermal: 0.21 in2/ft2 + 2%(1)]*
ar reinforcement provided: See [APP-GW-GLR-602, Section 4.]*
s:
[The Tier 2* designation for Plate thickness required requires NRC approval if this value is exceeded as a result of design changes or detail design adjustments identified during preparation of fabrication or construction drawings or instructions.]*
Load cases 23 and 41 are the two governing load combinations for element 23752 that include thermal and seismic loads combined numerically. W1 designates the winter conditions with the spent fuel pool at the normal operating temperature limit.
W2 designates the winter conditions with the spent fuel pool and fuel transfer canal at the normal operating temperature limit.
Es is SRSS (member forces are positive) of the SSE loads. E's is Es with all member forces except axial forces (TX, TY) reversed to negative.
Staff approval is required prior to implementing a change in this information.
3H-46 Revision 1
Provided Stress Required (Minimum) Reinforcement Critical Sections Component in2/ft in2/ft Ratio nical Roof Steel Beams]* (1) Axial + Bending - [Radial Beams 1.33 Shear - W36 X 393]* 8.33 nical Roof Near Tension Radial 1.80 [1.96]* 1.09 g]* Hoop 4.31 [4.68]* 1.09 uckle Region]* Vertical 1.37 [1.72]* 1.25 Radial 1.52 [2.23]* 1.47 Hoop 1.37 [3.12]* 2.28 mpression Ring]* Vertical 1.04 [1.48]* 1.42 Radial 3.09 [4.42]* 1.43 Hoop 2.14 [3.12]* 1.45 Steel beams are not considered as reinforcement for the reinforced concrete roof. Ratio for conical roof steel beams is based on demand and allowable stresses in psi.
Staff approval is required prior to implementing a change in this information.
3H-47 Revision 1
Figure 3H.2-1
[General Layout of Auxiliary Building]*
Staff approval is required prior to implementing a change in this information.
3H-48 Revision 1
Figure 3H.5-1 (Sheet 1 of 3)
[Nuclear Island Critical Sections Plan at El. 135-3]*
Staff approval is required prior to implementing a change in this information.
3H-49 Revision 1
Figure 3H.5-1 (Sheet 2 of 3)
[Nuclear Island Critical Sections Plan at El. 180-0]*
Staff approval is required prior to implementing a change in this information.
3H-50 Revision 1
WLS 1&2 - UFSAR Security-Related Information, Withhold Under 10 CFR 2.390d Figure 3H.5-1 (Sheet 3 of 3)
[Nuclear Island Critical Sections Section A-A]*
- NRC Staff approval is required prior to implementing a change in this information.
3H-51 Revision 1
Figure 3H.5-2 (Sheet 1 of 3)
[Wall on Column Line 1]*
Staff approval is required prior to implementing a change in this information.
3H-52 Revision 1
Figure 3H.5-2 (Sheet 2 of 3)
[Wall on Column Line 7.3]*
Staff approval is required prior to implementing a change in this information.
3H-53 Revision 1
Figure 3H.5-2 (Sheet 3 of 3)
[Wall on Column Line L]*
Staff approval is required prior to implementing a change in this information.
3H-54 Revision 1
Figure 3H.5-3
[Typical Reinforcement in Wall on Column Line 1]*
Staff approval is required prior to implementing a change in this information.
3H-55 Revision 1
Figure 3H.5-4
[Typical Reinforcement in Wall 7.3]*
Staff approval is required prior to implementing a change in this information.
3H-56 Revision 1
Figure 3H.5-5 (Sheet 1 of 3)
[Concrete Reinforcement in Wall 11]*
Staff approval is required prior to implementing a change in this information.
3H-57 Revision 1
Figure 3H.5-5 (Sheet 2 of 3)
[Concrete Reinforcement Layers in Wall 11 (Looking East)]*
Staff approval is required prior to implementing a change in this information.
3H-58 Revision 1
Figure 3H.5-5 (Sheet 3 of 3)
[Wall 11 at Main Steamline Anchor Section A-A]*
Staff approval is required prior to implementing a change in this information.
3H-59 Revision 1
Figure 3H.5-6
[Auxiliary Building Typical Composite Floor]*
Staff approval is required prior to implementing a change in this information.
3H-60 Revision 1
WLS 1&2 - UFSAR Figure 3H.5-7
[Typical Reinforcement and Connection to Shield Building]*
- NRC Staff approval is required prior to implementing a change in this information.
3H-61 Revision 1
WLS 1&2 - UFSAR Figure 3H.5-8
[Auxiliary Building Operations Work Area (Tagging Room) Ceiling]*
- NRC Staff approval is required prior to implementing a change in this information.
3H-62 Revision 1
Figure 3H.5-9 (Sheet 1 of 3)
[Auxiliary Building Finned Floor]*
Staff approval is required prior to implementing a change in this information.
3H-63 Revision 1
- 8@12 Construction Joint
- 7@12 65 135-3 FINNED FLOOR 24 Mechanical Splices 9
65 1/2 PL 1 Gap 1/2 PL @ 9
- 8@12 S N Wall 11 Figure 3H.5-9 (Sheet 2 of 3)
[Auxiliary Building Finned Floor]*
Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-9 (Sheet 3 of 3)
[Auxiliary Building Finned Floor]*
Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-10
[Spent Fuel Pool Wall Divider Wall Element Locations]*
Staff approval is required prior to implementing a change in this information.
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WLS 1&2 - UFSAR Security-Related Information, Withhold Under 10 CFR 2.390d Figure 3H.5-11 (Sheet 1 of 6)
[Design of Shield Building: Roof and Air Inlets]*
- NRC Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-11 (Sheet 2 of 6)
[Design of Shield Building: Concrete Detail at Tension Ring]*
Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-11 (Sheet 3 of 6)
[Design of Shield Building: Roof/Air Inlet Interface]*
Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-11 (Sheet 4 of 6)
[Design of Shield Building at Air Inlets]*
Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-11 (Sheet 5 of 6)
[Design of Shield Building: Tank/Roof Interface Reinforcement]*
Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-11 (Sheet 6 of 6)
Design of Shield Building: Tank/Compression Ring Roof Interface Reinforcement 3H-72 Revision 1
Figure 3H.5-12
[Typical Reinforcement in Wall L]*
Staff approval is required prior to implementing a change in this information.
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Figure 3H.5-13 Enhanced Shield Building Wall Panel Layout 3H-74 Revision 1
Figure 3H.5-14 Elevation View of Tension Ring and Air Inlets 3H-75 Revision 1
Figure 3H.5-15 Shield Building Tension Ring 3H-76 Revision 1
For additional information, see Figure 6 of APP-GW-GLR-602 (Reference 1).
Figure 3H.5-16 (Sheet 1 of 2)
[Design of Shield Building: Surface Plates on Cylindrical Section - Developed View 90-270 Degrees]*
Staff approval is required prior to implementing a change in this information.
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Security-Related Information, Withhold Under 10 CFR 2.390d For additional information, see Figure 6 of APP-GW-GLR-602 (Reference 1).
Figure 3H.5-16 (Sheet 2 of 2)
[Design of Shield Building: Surface Plates on Cylindrical Section - Developed View 270-90 Degrees]*
Staff approval is required prior to implementing a change in this information.
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seismic analysis and design of the AP1000 plant is based on the Certified Seismic Design ponse Spectra (CSDRS) shown in Subsection 3.7.1.1. These spectra are based on Regulatory de 1.60 with an increase in the 25 hertz region. Ground Motion Response Spectra (GMRS) for e Central and Eastern United States rock sites show higher amplitude at high frequency than the DRS. Evaluations are described in this appendix for an envelope response spectra with high uency for the seismic input. The resulting spectra of this site are shown in Figure 3I.1-1 and re 3I.1-2 and compare this hard rock high frequency (HRHF) envelope response spectra at the dation level against the AP1000 CSDRS for both the horizontal and vertical directions for 5%
ping. The HRHF envelope response spectra exceed the CSDRS for frequencies above about Hz.
h frequency seismic input is generally considered to be non-damaging as described in erence 1. The evaluation of the AP1000 nuclear island for the high frequency input is based on analysis of a limited sample of structures, components, supports, and piping to demonstrate that high frequency seismic response is non-damaging. The evaluation includes building structures, tor pressure vessel and internals, primary component supports, primary loop nozzles, piping, equipment.
appendix describes the methodology and criteria used in the evaluation to confirm that the high uency input is not damaging to equipment and structures qualified by analysis for the AP1000 DRS. It provides supplemental criteria for selection and testing of equipment whose function might ensitive to high frequency. The results of the high frequency evaluation demonstrating that the 000 plant is qualified for this type of input are documented in a technical report (Reference 2).
report will provide a summary of the analysis and test results.
nuclear island foundation input response spectra (NI FIRS) for Lee Nuclear Station, the envelope e GMRS (Unit 2 FIRS) and the Unit 1 FIRS (Subsection 3.7.1.1.1), are slightly above the 000 HRHF spectra, but the spectra are very similar. Figures 3I.1-201 and 3I.1-202 compare the IRS to the AP1000 CSDRS and the AP1000 HRHF spectra for the horizontal and vertical ctions for 5% damping. The NI FIRS exceeds the AP1000 CSDRS for frequencies above roximately 14 Hz and the AP1000 HRHF spectra above approximately 3 Hz.
ause the NI FIRS are not enveloped by the AP1000 HRHF spectra, a site-specific analysis is ormed to evaluate and justify exceedances. Technical report WLG-GW-GLR-815 ference 201) provides a summary of those evaluations and results. This report presents tructure response spectra throughout the Nuclear Island resulting from the site-specific input.
se in-structure response spectra were investigated and all exceedances of the CSDRS or HRHF ctra were identified. Three instances of largest exceedances were noted, and these three ances were investigated as bounding conditions and justified by further evaluations.
High Frequency Seismic Input sented in Figures 3I.1-1 and 3I.1-2 is a comparison of the horizontal and vertical HRHF envelope onse spectra and the AP1000 CSDRS. The HRHF envelope response spectra presented are ulated at foundation level (39.5' below grade), at the upper most competent material and treated n outcrop for calculation purposes.
each direction, the HRHF envelope response spectra exceed the design spectra in higher uencies (greater than 15 Hz horizontal and 20 Hz vertical). The spectra are used for the HRHF 3I-1 Revision 1
se HRHF envelope response spectra are further limited in that the shear wave velocity limitation efined at the bottom of the basemat equal to or higher than 7,500 fps, while maintaining a shear e velocity equal to or above 8,000 fps at the lower depths.
res 3I.1-201 and 3I.1-202 present a comparison of the horizontal and vertical (respectively) Lee lear Station NI FIRS to the AP1000 CSDRS and the AP1000 HRHF. The NI FIRS are calculated undation level (39.5' below grade), at the upper most competent material and treated as an rop for calculation purposes.
each direction, the NI FIRS exceeds the CSDRS in higher frequencies (greater than 14 Hz zontal and 16 Hz vertical) and the AP1000 HRHF spectra at frequencies greater than 3 Hz in both horizontal and vertical directions.
NI Models Used To Develop High Frequency Response NI20 nuclear island model described in Appendix 3G is analyzed in ACS SASSI using the HRHF histories applied at foundation level to obtain the motion at the base.
odal analysis of the NI05 model for both the auxiliary and shield buildings and containment rnal structure (CIS) has been performed for each of these regions. Specific areas within each wall oor where out-of-plane modes, which may respond to either CSDRS or HRHF input (including ctures with modes less than 33 Hz and between 33 Hz to 50 Hz), have been identified. The ey reveals that some regions, typically in the middle of a floor or wall, exhibit amplified behavior pared to the critical nodes at the corner and edge building locations. The amplified FRS for these ons is generated in addition to the typical set of critical nodes for building analysis by a single history analysis of the NI05 building model subject to the HRHF time history input. Seismic onse spectra for each of the flexible nodes are considered when selecting the pre-existing up spectra, which is the envelope of the entire floor in that area.
luation of incoherent HRHF spectra has been performed. The CSDRS and HRHF seismic onses were compared with coherent and incoherent considerations at a number of locations in nuclear island. There are some exceedances, mostly above the 15 hertz region, and these are cal of the plant comparative responses. The steel containment vessel (SCV) was excluded from evaluation because the HRHF spectra at the base of the SCV are enveloped by the AP1000 DRS spectra at the base of the SCV.
ctures designed to the CSDRS input are adequately designed for the HRHF input because the HF coherent results are enveloped by the CSDRS results.
NI20u nuclear island model (Reference 201) is analyzed in ACS SASSI using the Lee Nuclear ion NI FIRS time histories (Subsection 3.7.2.1.2) applied at foundation level to obtain the motion e base.
NI20u model used in the Lee Nuclear Station site-specific analysis was updated to incorporate ign changes from detailed design finalization of the AP1000 standard plant (no impact from ign changes to licensing basis as defined in AP1000 DCD Rev 19) and to improve the match ween the NI20u model and the more realistic NI10 model used to design and qualify the AP1000 dard plant for the CSDRS.
luation of incoherent NI FIRS has been performed. In-structure response spectra for the AP1000 DRS, incoherent HRHF spectra and the incoherent NI FIRS were compared at a number of 3I-2 Revision 1
Evaluation Methodology demonstration that the AP1000 nuclear power plant is qualified for the high frequency seismic onse does not require the analysis of the total plant. The evaluations made are of representative ems, structures, and components, selected by screening, as potentially sensitive to high uency input in locations where there were exceedances in the high frequency region.
eptability of this sample is considered sufficient to demonstrate that the AP1000 is qualified.
high frequency seismic analyses that are performed use time history or broadened response ctra. The analysis is not performed using the combination spectra of the CSDRS and the HRHF elope response spectra. Separate analyses with each spectra are used.
high frequency seismic analyses used the soil-structure interaction code ACS SASSI. The lts presented in this report are based on the stochastic (multiple, statistical analyses) seismic herent soil-structure interaction analysis approach referred herein as the simulation approach.
evaluations performed assess the ability of the system, structure, or component to maintain its ty function.
plementary analyses are performed as needed to show that high frequency floor response ctra exceedances are not damaging. These analyses can include: gap nonlinearities; material astic behavior; multi point response spectra analyses where the high frequency response excites cal part of the system. Tests on equipment are specified as needed where function cannot be onstrated by analysis, or analysis is not appropriate.
General Selection Screening Criteria following general screening criteria are used to identify representative AP1000 systems, ctures, and components (SSCs) for the samples to be evaluated to demonstrate acceptability of AP1000 nuclear power plant for the high frequency motion.
Select systems, structures, and components based on their importance to safety. This includes the review of component safety function for the SSE event and its potential failure modes due to an SSE. Those components whose failure modes would result in safe shutdown are excluded.
Select systems, structures, and components that are located in areas of the plant that experience large high frequency seismic response.
Select systems, structures, and components that have significant modal response within the region of high frequency amplification. Significance is defined by such items as modal mass; participation factor, stress and/or deflection.
Select systems, structures, and components that have significant stress as compared to allowable when considering load combinations that include seismic.
Evaluation is section the portions of structures, the components, and the systems that are evaluated for the frequency seismic response are identified. The sample to be evaluated, based on the screening ria applicable to the SSCs consists of the following:
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- Shield Building - 8 locations
- CIS - 2 locations Primary Coolant Loop
- Reactor Vessel and Internals
- Primary Component Supports
- Reactor Coolant Loop Primary Equipment Nozzles Piping Systems - ASME Class 1, 2, and 3 piping systems will be evaluated for the HRHF GMRS. This evaluation is within the scope of the piping DAC (see COL Information Item 3.9-7).
Electro-Mechanical Equipment - Equipment that is potentially sensitive to high frequency input (see Table 3I.6-1) se structures, systems, and equipment are discussed in more detail in the sections that follow.
described in Lee Nuclear Station site-specific Technical Report WLG-GW-GLR-815 ference 201), all exceedances of the in-structure response spectra resulting from the Lee Nuclear ion NI FIRS input were identified. Three instances of largest in-structure response spectra eedances were investigated as bounding conditions and justified by further evaluation. Therefore, sample of structures, systems and components selected for evaluation remains unchanged.
.1 Building Structures ntaining the NI buildings structural integrity is important to the safety of the plant. Representative ions of the buildings that are evaluated for the effect of high frequency input are selected based hose areas that can experience high seismic shear and moment loads due to the seismic event.
as chosen are at the base of the shield building, in the vicinity of auxiliary building floors that have amental frequencies in the high frequency region, and the corners of the auxiliary building. Three tions are selected on the auxiliary building that reflect the bottom of a wall where the shear and ment would be large, a wall in the vicinity of a floor that is influenced by high frequency response, a corner intersection of walls. Eight locations are evaluated on the shield building. Four at ation 107' and four at elevation 211'. These locations are located on the east, west, north and th sides. The south-west wall of the refueling canal is evaluated since it is a representative wall on refueling canal. The CA02 wall in the CIS building is evaluated since it is a representative wall ociated with the IRWST.
evaluation consists of a comparison of the loads from the high frequency input to those obtained the AP1000 design spectra, shown in Figures 3I.1-1 and 3I.1-2, for these representative ding structures. The NI building structures are considered qualified for the high frequency input if seismic loads from the Regulatory Guide 1.60 (modified) envelope those from the high frequency
- t. If there is any exceedance, this is evaluated further to confirm that the existing design is quate.
d comparisons for the building structures evaluated show that the seismic loads resulting from CSDRS input motion are greater than the seismic loads generated from the NI FIRS ference 201).
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ndary. Therefore, it is chosen for evaluation. The components evaluated are as follows:
Reactor vessel and internals Reactor vessel supports Steam generator supports Reactor coolant loop primary equipment nozzles reactor vessel and internals are selected since they are important to safety and their analysis is esentative of major primary components. The building structure below the reactor vessel ports is fairly stiff and there may be significant vertical amplification at the supports of the reactor sure vessel. Further, reactor vessel internals have relatively complex structural systems uding gap nonlinearities and sliding elements. Also, they may be sensitive to high frequency input ummarized below:
Vertical and horizontal modes of the upper internals and the reactor vessel modes are in the relatively high frequency range.
Additional high frequencies are associated with nonlinear impact evaluation consists of a comparison of the loads from the high frequency input to those obtained the Regulatory Guide 1.60 (modified) input. Qualification is shown for the high frequency input if seismic loads from the Regulatory Guide 1.60 (modified) envelope those from the high frequency
- t. If there is exceedance, then comparison is made for the combination of the seismic with the ign basis pipe break loads and steady state loads. Qualification is then shown if the high uency loads are relatively insignificant compared to the other loads, or there are no required ign changes.
ntaining the integrity of the reactor vessel and steam generator supports is important to erving the primary component safety function. They are representative of supports on ponents, and see high loads.
reactor coolant loop nozzles at the cold and hot leg interfaces of the reactor pressure vessel, tor coolant pumps, and steam generators are important to include in the evaluation since these critical areas of components.
evaluation of the primary component supports and reactor coolant loop nozzles consists of a parison of the loads from the high frequency input to those obtained from the Regulatory de 1.60 (modified) input. These items are considered qualified for the high frequency input if the mic loads from the Regulatory Guide 1.60 (modified) envelope those from the high frequency
- t. If there is any exceedance, then an evaluation is made combining the high frequency loads the other load components (e.g., thermal, pressure, dead) and a comparison made to the design
- s. If the design loads envelope the load combinations that include the high frequency seismic t, then the nozzles and supports are considered qualified for the high frequency input.
d comparisons for the primary component supports and nozzles evaluated show that the seismic s resulting from the CSDRS input motion are greater than the seismic loads generated from the IRS (Reference 201).
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in the scope of the piping DAC (see COL Information Item 3.9-7).
ME Class 1, 2, and 3 piping packages were reviewed along with local input seismic response ctra for susceptibility to excitation from high frequency seismic input motion. Since the in-structure r response spectra (FRS) generated from the Lee Nuclear Station NI FIRS are enveloped pletely by either by the FRS generated from the CSDRS or HRHF spectra in most locations, all of piping analyses do not need to be redone for the NI FIRS.
ee piping packages, ADS 4th Stage East Compartment and Passive RHR Supply, Pressurizer ge Line, and SFS from Auxiliary Building Area 4 SCV to Auxiliary Building Area 6 SFS Pumps S Aux. Building 4 to 6) were chosen for evaluation (Reference 201). These packages are esentative of all safety class piping in Lee Nuclear Station because they are the most susceptible xcitation from high frequency seismic input motion.
stress results of the sample piping analysis packages show that the AP1000 HRHF stresses e greater than the NI FIRS stresses for all nodes in the ADS 4th Stage and SFS Aux. Building 4 piping packages and only slightly less in the Pressurizer Surge Line piping package. Stress parison results show that AP1000 CSDRS stresses are greater than the NI FIRS stresses at all es in all three piping packages except for one node in the SFS Aux. Building 4 to 6 piping kage where there was a slight NI FIRS exceedance. At this one point, the stresses resulting from NI FIRS were less than those from the HRHF spectra. Therefore, the design practices for dard plant AP1000 piping systems have considered cases that envelope the Lee site-specific uirements.
stresses due to the Lee Nuclear Station NI FIRS input are bounded by design basis analysis lts. The same applies to all of the analyzed piping supports. As a result, the effect of the NI FIRS t on safety class piping is found to be nondamaging (Reference 201).
.4 Electrical and Electro-Mechanical Equipment groups of safety-related equipment considered for evaluation are those that may be sensitive to high frequency input. This includes those cabinet-mounted equipment, field sensors, and urtenants that may be sensitive to high frequency seismic inputs identified in Table 3I.6-1.
ple safety-related cabinets have been identified that are typically sensitive to seismic input.
luations were performed to verify these cabinets do not have excessive seismic excitation on r mounted equipment, the cabinet designs do not require changes due to the high frequency t, and the cabinets will maintain their structural integrity during the high frequency input. Time ory analyses of these cabinets were performed for both the Regulatory Guide 1.60 (modified) and high frequency inputs so that comparisons can be made to their seismic response from both mic inputs. This analytical study reported in APP-GW-GLR-115 (Reference 2) concluded that ty-related equipment may be screened.
AP1000 HRHF screening program for determination and evaluation of potential high frequency sitive equipment is in compliance with the NRC requirements in Section 4.0, Identification and luation of HF Sensitive Mechanical and Electrical Equipment/Components, of COL/DC-ISG-1 ference 3). The AP1000 HRHF screening program is also consistent with the guidelines eloped as part of an industry review document in the EPRI White Paper, Seismic Screening of ponents Sensitive to High Frequency Vibratory Motions (Reference 4), transmitted to the NRC une 28, 2007, for determining the safety-related equipment and components that may be HRHF sitive, and screening procedures to ensure that any safety-related equipment and components 3I-6 Revision 1
ponents.
AP1000 HRHF screening program is based on an HF evaluation study reported in
-GW-GLR-115 (Reference 2). The HF evaluation study concluded that AP1000 In-Structure ponse Spectra (ISRS) developed from the AP1000 CSDRS would, in the majority of cases, duce equipment stress results of the same magnitude or higher than the stress results produced HRHF seismic excitation. The exception to this condition is when the dominant natural uency of the equipment is in the HRHF exceedance range and there can be significantly more onse because the frequency coincides with the input driving force. Under this condition, forces/
sses generated in the equipment could be due to the acceleration exceedance; therefore, the ipment will be subjected to HRHF seismic evaluation/testing to screen out equipment by verifying erformance and acceptability under HRHF excitation. Review of seismic test data for electrical microprocessor based cabinets performed to generic and high frequency excitation concluded seismic testing that peaks in the lower frequency range will produce larger displacements and cities, and will result in higher stresses in the equipment.
goal of the AP1000 HRHF screening program is to identify the potential safety-related equipment components that have the potential to be HRHF-sensitive and show them to be acceptable for r specific application (screened-out). The AP1000 HRHF screening program is a two step ess. The first step is an HRHF susceptibility review to identify potential high frequency sensitive ty-related equipment. The second step is the screened-out equipment process to demonstrate its eptability for the HRHF seismic excitation. Evaluation of screened-in equipment as defined in L/DC-ISG-1 (Reference 3) is not performed because all safety-related equipment that is ened-in will be eliminated or shown to be acceptable through a design change process.
the AP1000 HRHF screening program, the following conditions must exist:
Plant-specific HRHF GMRS exceeds the AP1000 CSDRS in the high frequency range at 5% critical damping.
Safety-related equipment has potential failure modes involving change of state, chatter, signal change/drift, and connection problems.
le 3I.6-2 is a list of potential HRHF-sensitive AP1000 safety-related equipment developed based able 3.11-1 of Section 3.11, Environmental Qualification of Mechanical and Electrical ipment. The equipment in Table 3.2-3 of Section 3.2, AP1000 Classification of Mechanical and d Systems, Components, and Equipment, and Table 3I.6-3 is not HRHF-sensitive. The structural grity and operability of equipment in Table 3.2-3 and Table 3I.6-3 will not be impacted by the high uency excitation.
HRHF susceptibility review of AP1000 safety-related equipment is not performed for potential re modes associated with mounting, connections, fasteners, joints, and structural interface.
se potential failure modes are addressed through the seismic qualification of the safety-related ipment to the AP1000 ISRS testing performed in compliance with IEEE Standard 344-1987. The 000 ISRS qualification testing generates higher displacements and velocities than those lting from HRHF seismic excitation since the AP1000 ISRS is controlled by the lower frequency ge. The higher displacement, velocities, and accelerations will detect these equipment structural re modes if they exist.
cations where HRHF response spectra show exceedance of the CSDRS and there is a likelihood quipment damage, further evaluations would be performed to verify that the existing qualification 3I-7 Revision 1
Safety-related equipment must have modes or natural frequencies in the range of interest.
Evaluation will apply the same acceptance criteria and methodologies used in CSDRS qualification.
emonstrate acceptability for both CSDRS and HRHF testing, the test response spectra must elop the CSDRS and HRHF spectra, respectively, with margin over the frequency range of rest in compliance with IEEE Standard 344-1987. In the event that the CSDRS and/or HRHF onse spectra would be revised after the qualification program has been completed, a nciliation effort would be performed to verify that the CSDRS and HRHF testing is still valid. The nciliation effort may result in requalification activities and qualification documentation revisions.
h Frequency Screening Process - Step 1 potential failure modes of high frequency sensitive component types and assemblies are ortant considerations in the high frequency program. The following are potential failure modes of frequency sensitive components/equipment:
Inadvertent change of state Chatter Change in accuracy and drift in output signal or set-point Electrical connection failure or intermediacy (e.g., poor quality solder joints)
Mechanical connection failure Mechanical misalignment/binding (e.g., latches, plungers)
Fatigue failure (e.g., solder joints, ceramics, self-taping screws, spot welds)
Improperly and unrestrained mounted components Inadequately secured/locked mechanical fasteners and connections ponents and equipment determined to have potential failure modes involve change of state, tter, signal change/drift, and connection problems will be demonstrated to be acceptable through performance of a supplemental high frequency screening test. Those high frequency sensitive ponents having failure modes associated with mounting, connections and fasteners, joints, and rface are considered to be acceptable as a result of the AP1000 ISRS qualification testing per E Standard 344-1987 and/or require quality assurance inspection and process/design controls.
h Frequency Screening Process - Step 2 HRHF susceptibility review is to verify that the subject equipment is capable of performing its ty-related function under HRHF seismic excitation. All AP1000 safety-related equipment will be lified to the AP1000 ISRS, and the dominant natural frequencies of the equipment will be rmined. The EPRI White Paper (Reference 4) identifies the following three evaluation methods emonstrate that potential HRHF-sensitive safety-related equipment is not HRHF vulnerable:
Existing seismic qualification test data for potential high frequency sensitive equipment should be reviewed for applicability and adequacy of the test method to demonstrate sufficient high frequency content.
Systems/circuits containing potentially sensitive items should be reviewed for inappropriate/
unacceptable system actions due to assumed change of state, contact chatter/intermittency, set point drifts, or loss of calibration.
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first and third evaluation methods are part of the AP1000 HRHF screening program and are her detailed below. The AP1000 HRHF seismic screening evaluation will employ the AP1000 HF SSE response spectra as input in verifying potential HF sensitive safety-related equipment is vulnerable to HRHF seismic excitation. Additional seismic test margin will be introduced into the HF seismic screening evaluation as needed.
hod 1: Review of Seismic Test Data ilable seismic test data can be used for AP1000 HRHF plant applications when:
Seismic qualification testing performed on potential HRHF-sensitive safety-related equipment meets as a minimum the AP1000 ISRS in compliance with IEEE Standard 344-1987.
Safe shutdown earthquake (SSE) test had sufficient energy content in the HRHF region to verify that the safety-related equipment is not vulnerable to HRHF seismic excitation.
additional seismic testing is required for safety-related equipment previously tested and whose lification level envelops the HRHF required response spectra (RRS).
E Standard 344-1987 provides guidance to ensure that the seismic test input is generated and in pliance with the frequency range of interest. To demonstrate acceptability for frequency content, necessary to show that the frequency content of the test waveform is at least as broad as the uency content of the amplified region of the RRS except at the low frequencies where non-eloping is permitted under certain conditions (refer to IEEE Standard 344-1987 clauses 7.6.3.1(10) and 7.6.3.1(13)). An evaluation of the test input waveform should be ducted per IEEE Standard 344-1987 Annex B to verify the test data has sufficient content over the uency range of interest throughout the input time history. If an evaluation of the test input is ormed, and the data demonstrates sufficient frequency content in the high frequency range ughout the time history, then the data is acceptable.
hod 3: HRHF Screening Test HRHF screening test is a supplemental test to the required seismic qualification methods ormed in accordance with IEEE Standard 344-1987 for those plants that have high frequency eedance of the AP1000 CSDRS. The purpose of the HRHF screening test is to demonstrate that potential HRHF-sensitive safety-related equipment will perform its safety-related function as uired under HRHF seismic excitation. The HRHF screening test is performed in conjunction with AP1000 ISRS seismic qualification testing, or it is performed as a supplemental test after pletion of the AP1000 ISRS seismic qualification testing. The AP1000 ISRS and HRHF test input histories have 30-second durations with frequency content up to the cutoff frequency developed ccordance with subclause 7.6.3 (Multiple-Frequency Tests) and Annex B (Frequency Content and ionarity) of IEEE Standard 344-1987. During the AP1000 ISRS and HRHF testing, the equipment be functional and monitored to verify the safety-related function was demonstrated. Screening ing will be performed using HRHF response spectra as defined in the EPRI White Paper ference 4) when AP1000 HRHF inputs are not available. The HRHF response spectra will be erated based on the 5g and 15g peak spectral acceleration at 5% critical damping in the 25 Hz to Hz frequency range. If the HRHF screening test cannot demonstrate the equipment to be eptable, then the safety-related equipment will be removed or modified and additional testing or fication will be required.
AP1000 safety-related equipment will be seismic qualified to the AP1000 ISRS associated with mounting location of the equipment as a minimum. Seismic qualification testing will consist of five 3I-9 Revision 1
eedance area is adequately addressed by performing five OBE (one-half the SSE) and a imum of one SSE seismic test runs in compliance with IEEE Standard 344-1987 prior to orming the supplemental HRHF screening test. Additional OBE testing in the high frequency eedance range is adequately addressed by the demonstration that the peak stress cycles uired for five one-half SSE events using the AP1000 HRHF ISRS are equivalent to or enveloped he peak stress cycles resulting from five one-half SSE events and one full SSE event using the 000 CSD ISRS.
test results of AP1000 seismic qualification programs with multiple operational states (for mple, relays have three possible operational states: de-energized, energized, and change of e) will be used to determine the most sensitive equipment electrical operational state. The HRHF run is performed on the equipment in its most sensitive electrical operational state to onstrate its safety-related function under HRHF seismic excitation. If this is not possible, itional HRHF screening tests will be performed as needed to address the other most sensitive trical operation states.
emonstrate acceptability, the test response spectra (TRS) for high frequency sensitive equipment ured for Lee Nuclear Station will have to bound the required response spectra (RRS) of the 000 CSDRS, AP1000 HRHF spectra, and the NI FIRS generated in-structure response spectra.
hown in the Lee Nuclear Station site-specific Technical Report WLG-GW-GLR-815 ference 201), very little if any of the AP1000 equipment will need to be re-qualified for the Lee lear Station high frequency seismic motion considering margins in the TRS currently being used ualify AP1000 high frequency sensitive equipment. However, per the licensing commitment in section 3.7.2.15, Duke Energy will ensure that all seismic qualification testing for safety-related ipment required per this Appendix appropriately envelopes the Lee Nuclear Station site-specific mic requirements, in addition to the CSDRS and HRHF RRS.
References EPRI Draft White Paper, Considerations for NPP Equipment and Structures Subjected to Response Levels Caused by High Frequency Ground Motions, Transmitted to NRC March 19, 2007.
APP-GW-GLR-115, Effect of High Frequency Seismic Content on SSCs, Westinghouse Electric Company LLC.
COL/DC-ISG-1, Interim Staff Guidance on Seismic Issues of High Frequency Ground Motion, May 19, 2008.
EPRI White Paper, Seismic Screening of Components Sensitive to High Frequency Vibratory Motions, June 2007.
Letter, R. Sisk (Westinghouse) to NRC, AP1000 Response to Request for Additional Information (SRP3.10), DCP/NRC2280, October 17, 2008.
. Westinghouse Electric Company, LLC, "Effect of William S. Lee Site Specific Seismic Requirements on AP1000 SSCs," WLG-GW-GLR-815, Revision 0, January 17, 2014.
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quipment or components with moving parts and required to perform a switching function during the seismic vent (e.g., circuit breakers, contactors, auxiliary switches, molded case circuit breakers, motor control center tarters, and pneumatic control assemblies)
Components with moving parts that may bounce or chatter such as relays and actuation devices (e.g., shunt trips)
Unrestrained components otentiometers rocess switches and sensors (e.g., pressure/differential pressure, temperature, level, limit/position, and flow)
Components with accuracy requirements that may drift due to seismic loading nterfaces such as secondary contacts Connectors and connections (including circuit board connections for digital and analog equipment) 3I-11 Revision 1
AP1000 Safety-Related electrical and Electro-mechanical Equipment AP1000 Description Tag Number teries A 125V 60 Cell Battery 1A IDSA-DB-1A A 125V 60 Cell Battery 1B IDSA-DB-1B B 125V 60 Cell Battery 1A IDSB-DB-1A B 125V 60 Cell Battery 1B IDSB-DB-1B B 125V 60 Cell Battery 2A IDSB-DB-2A B 125V 60 Cell Battery 2B IDSB-DB-2B C 125V 60 Cell Battery 1A IDSC-DB-1A C 125V 60 Cell Battery 1B IDSC-DB-1B C 125V 60 Cell Battery 2A IDSC-DB-2A C 125V 60 Cell Battery 2B IDSC-DB-2B D 125V 60 Cell Battery 1A IDSD-DB-1A D 125V 60 Cell Battery 1B IDSD-DB-1B re 125V 60 Cell Battery 1A IDSS-DB-1A re 125V 60 Cell Battery 1B IDSS-DB-1B tery Chargers A Battery Charger IDSA-DC-1 B Battery Charger IDSB-DC-1 B Battery Charger 2 IDSB-DC-2 C Battery Charger 1 IDSC-DC-1 C Battery Charger 2 IDSC-DC-2 D Battery Charger IDSD-DC-1 re Battery Charger IDSS-DC-1 3I-12 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number tribution Panels A 250 Vdc Dist Panel IDSA-DD-1 B 250 Vdc Dist Panel IDSB-DD-1 C 250 Vdc Dist Panel IDSC-DD-1 D 250 Vdc Dist Panel IDSD-DD-1 A 120 Vac Dist Panel 1 IDSA-EA-1 A 120 Vac Dist Panel 2 IDSA-EA-2 B 120 Vac Dist Panel 1 IDSB-EA-1 B 120 Vac Dist Panel 2 IDSB-EA-2 B 120 Vac Dist Panel 3 IDSB-EA-3 C 120 Vac Dist Panel 1 IDSC-EA-1 C 120 Vac Dist Panel 2 IDSC-EA-2 C 120 Vac Dist Panel 3 IDSC-EA-3 D 120 Vac Dist Panel 1 IDSD-EA-1 D 120 Vac Dist Panel 2 IDSD-EA-2 e Panels A Fuse Panel IDSA-EA-4 B Fuse Panel IDSB-EA-4 B Fuse Panel IDSB-EA-5 B Fuse Panel IDSB-EA-6 C Fuse Panel IDSC-EA-4 C Fuse Panel IDSC-EA-5 C Fuse Panel IDSC-EA-6 D Fuse Panel IDSD-EA-4 3I-13 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number nsfer Switches A Fused Transfer Switch Box 1 IDSA-DF-1 B Fused Transfer Switch Box 1 IDSB-DF-1 B Fused Transfer Switch Box 2 IDSB-DF-2 C Fused Transfer Switch Box 1 IDSC-DF-1 C Fused Transfer Switch Box 2 IDSC-DF-2 D Fused Transfer Switch Box 1 IDSD-DF-1 S Fused Transfer Switch Box 1 IDSS-DF-1 re Battery 125/250 Vdc Disconnect Switch IDSS-SW-1 S Spare Termination Box IDSS-DF-2 S Spare Termination Box IDSS-DF-3 S Spare Termination Box IDSS-DF-4 S Spare Termination Box IDSS-DF-5 S Spare Termination Box IDSS-DF-6 or Control Centers A 250 Vdc MCC IDSA-DK-1 B 250 Vdc MCC IDSB-DK-1 C 250 Vdc MCC IDSC-DK-1 D 250 Vdc MCC IDSD-DK-1 tchboards A 250 Vdc Switchboard 1 IDSA-DS-1 B 250 Vdc Switchboard 1 IDSB-DS-1 B 250 Vdc Switchboard 2 IDSB-DS-2 C 250 Vdc Switchboard 1 IDSC-DS-1 C 250 Vdc Switchboard 2 IDSC-DS-2 D 250 Vdc Switchboard 1 IDSD-DS-1 3I-14 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number nsformers A Regulating Transformer 1 IDSA-DT-1 B Regulating Transformer 1 IDSB-DT-1 C Regulating Transformer 1 IDSC-DT-1 D Regulating Transformer 1 IDSD-DT-1 erters A Inverter IDSA-DU-1 B Inverter 1 IDSB-DU-1 B Inverter 2 IDSB-DU-2 C Inverter 1 IDSC-DU-1 C Inverter 2 IDSC-DU-2 D Inverter IDSD-DU-1 tchgear P 1A 6900V Switchgear 31 ECS-ES-31 P 1A 6900V Switchgear 32 ECS-ES-32 P 2A 6900V Switchgear 51 ECS-ES-51 P 2A 6900V Switchgear 52 ECS-ES-52 P 1B 6900V Switchgear 41 ECS-ES-41 P 1B 6900V Switchgear 42 ECS-ES-42 P 2B 6900V Switchgear 61 ECS-ES-61 P 2B 6900V Switchgear 62 ECS-ES-62 ctor Trip Switchgear PMS-JD-RTSA01 ctor Trip Switchgear PMS-JD-RTSA02 ctor Trip Switchgear PMS-JD-RTSB01 ctor Trip Switchgear PMS-JD-RTSB02 3I-15 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number ctor Trip Switchgear PMS-JD-RTSC01 ctor Trip Switchgear PMS-JD-RTSC02 ctor Trip Switchgear PMS-JD-RTSD01 ctor Trip Switchgear PMS-JD-RTSD02 el Switches e Makeup Tank A Narrow Range PXS-JE-LS011A e Makeup Tank A Narrow Range PXS-JE-LS011B e Makeup Tank A Narrow Range PXS-JE-LS011C e Makeup Tank A Narrow Range PXS-JE-LS011D e Makeup Tank B Narrow Range PXS-JE-LS012A e Makeup Tank B Narrow Range PXS-JE-LS012B e Makeup Tank B Narrow Range PXS-JE-LS012C e Makeup Tank B Narrow Range PXS-JE-LS012D e Makeup Tank A Narrow Range PXS-JE-LS013A e Makeup Tank A Narrow Range PXS-JE-LS013B e Makeup Tank A Narrow Range PXS-JE-LS013C e Makeup Tank A Narrow Range PXS-JE-LS013D e Makeup Tank B Narrow Range PXS-JE-LS014A e Makeup Tank B Narrow Range PXS-JE-LS014B e Makeup Tank B Narrow Range PXS-JE-LS014C e Makeup Tank B Narrow Range PXS-JE-LS014D tainment Floodup Level PXS-JE-LS050 tainment Floodup Level PXS-JE-LS051 tainment Floodup Level PXS-JE-LS052 3I-16 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number tron Detectors rce Range Neutron Detector RXS-JE-NE001A rce Range Neutron Detector RXS-JE-NE001B rce Range Neutron Detector RXS-JE-NE001C rce Range Neutron Detector RXS-JE-NE001D rmediate Range Neutron Detector RXS-JE-NE002A rmediate Range Neutron Detector RXS-JE-NE002B rmediate Range Neutron Detector RXS-JE-NE002C rmediate Range Neutron Detector RXS-JE-NE002D er Range Neutron Detector (Lower) RXS-JE-NE003A er Range Neutron Detector (Lower) RXS-JE-NE003B er Range Neutron Detector (Lower) RXS-JE-NE003C er Range Neutron Detector (Lower) RXS-JE-NE003D er Range Neutron Detector (Upper) RXS-JE-NE004A er Range Neutron Detector (Upper) RXS-JE-NE004B er Range Neutron Detector (Upper) RXS-JE-NE004C er Range Neutron Detector (Upper) RXS-JE-NE004D iation Monitors tainment High Range Area Monitor PXS-JE-RE160 tainment High Range Area Monitor PXS-JE-RE161 tainment High Range Area Monitor PXS-JE-RE162 tainment High Range Area Monitor PXS-JE-RE163 trol Room Supply Air Area Monitor VBS-JE-RE001A trol Room Supply Air Area Monitor VBS-JE-RE001B 3I-17 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number ed Sensors P 1A Pump Speed RCS-JE-ST281 P 1B Pump Speed RCS-JE-ST282 P 2A Pump Speed RCS-JE-ST283 P 2B Pump Speed RCS-JE-ST284 nsmitters S Water Delivery Flow PCS-JE-FT001 S Water Delivery Flow PCS-JE-FT002 S Water Delivery Flow PCS-JE-FT003 S Water Delivery Flow PCS-JE-FT004 S Storage Tank Water Level PCS-JE-LT010 S Storage Tank Water Level PCS-JE-LT011 HR HX Flow PXS-JE-FT049A HR HX Flow PXS-JE-FT049B S Hot Leg 1 Flow RCS-JE-FT101A S Hot Leg 1 Flow RCS-JE-FT101B S Hot Leg 1 Flow RCS-JE-FT101C S Hot Leg 1 Flow RCS-JE-FT101D S Hot Leg 2 Flow RCS-JE-FT102A S Hot Leg 2 Flow RCS-JE-FT102B S Hot Leg 2 Flow RCS-JE-FT102C S Hot Leg 2 Flow RCS-JE-FT102D 1 Startup Feedwater Flow SGS-JE-FT055A 1 Startup Feedwater Flow SGS-JE-FT055B 2 Startup Feedwater Flow SGS-JE-FT-056A 3I-18 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number 2 Startup Feedwater Flow SGS-JE-FT056B R Air Delivery Line Flow Rate - A VES-JE-FT003A R Air Delivery Line Flow Rate - B VES-JE-FT003B nt Vent Flow VFS-JE-FT101 ST Level PXS-JE-LT045 ST Level PXS-JE-LT046 ST Level PXS-JE-LT047 ST Level PXS-JE-LT048 S Hot Leg Water Level RCS-JE-LT160A S Hot Leg Water Level RCS-JE-LT160B Level RCS-JE-LT195A Level RCS-JE-LT195B Level RCS-JE-LT195C Level RCS-JE-LT195D 1 Narrow Range Level SGS-JE-LT001 1 Narrow Range Level SGS-JE-LT002 1 Narrow Range Level SGS-JE-LT003 1 Narrow Range Level SGS-JE-LT004 2 Narrow Range Level SGS-JE-LT005 2 Narrow Range Level SGS-JE-LT006 2 Narrow Range Level SGS-JE-LT007 2 Narrow Range Level SGS-JE-LT008 1 Wide Range Level SGS-JE-LT011 1 Wide Range Level SGS-JE-LT012 1 Wide Range Level SGS-JE-LT015 3I-19 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number 1 Wide Range Level SGS-JE-LT016 2 Wide Range Level SGS-JE-LT013 2 Wide Range Level SGS-JE-LT014 2 Wide Range Level SGS-JE-LT017 2 Wide Range Level SGS-JE-LT018 nt Fuel Pool Level SFS-JE-LT019A nt Fuel Pool Level SFS-JE-LT019B nt Fuel Pool Level SFS-JE-LT019C Storage Tank Pressure - A VES-JE-PT001A Storage Tank Pressure - B VES-JE-PT001B tainment Pressure Normal Range PCS-JE-PT005 tainment Pressure Normal Range PCS-JE-PT006 tainment Pressure Normal Range PCS-JE-PT007 tainment Pressure Normal Range PCS-JE-PT008 tainment Pressure Extended Range PCS-JE-PT012 tainment Pressure Extended Range PCS-JE-PT013 tainment Pressure Extended Range PCS-JE-PT014 S Wide Range Pressure RCS-JE-PT140A S Wide Range Pressure RCS-JE-PT140B S Wide Range Pressure RCS-JE-PT140C S Wide Range Pressure RCS-JE-PT140D Pressure RCS-JE-PT191A Pressure RCS-JE-PT191B Pressure RCS-JE-PT191C Pressure RCS-JE-PT191D 3I-20 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number n Steam Line SG1 Pressure SGS-JE-PT030 n Steam Line SG1 Pressure SGS-JE-PT031 n Steam Line SG1 Pressure SGS-JE-PT032 n Steam Line SG1 Pressure SGS-JE-PT033 n Steam Line SG2 Pressure SGS-JE-PT034 n Steam Line SG2 Pressure SGS-JE-PT035 n Steam Line SG2 Pressure SGS-JE-PT036 n Steam Line SG2 Pressure SGS-JE-PT037 n Control Room Differential Pressure VES-JE-PDT004A n Control Room Differential Pressure VES-JE-PDT004B tection and Safety Monitoring Systems tection and Safety Monitoring System Cabinets Multiple R/RSW Transfer Switch Panel A PMS-JW-004A R/RSW Transfer Switch Panel B PMS-JW-004B R/RSW Transfer Switch Panel C PMS-JW-004C R/RSW Transfer Switch Panel D PMS-JW-004D rce Range Neutron Flux Preamplifier Panel A PMS-JW-005A rce Range Neutron Flux Preamplifier Panel B PMS-JW-005B rce Range Neutron Flux Preamplifier Panel C PMS-JW-005C rce Range Neutron Flux Preamplifier Panel D PMS-JW-005D rmediate Range Neutron Flux Preamplifier Panel A PMS-JW-006A rmediate Range Neutron Flux Preamplifier Panel B PMS-JW-006B rmediate Range Neutron Flux Preamplifier Panel C PMS-JW-006C 3I-21 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number rmediate Range Neutron Flux Preamplifier Panel D PMS-JW-006D er Range Neutron Flux High Voltage Distribution Box A PMS-JW-007A er Range Neutron Flux High Voltage Distribution Box B PMS-JW-007B er Range Neutron Flux High Voltage Distribution Box C PMS-JW-007C er Range Neutron Flux High Voltage Distribution Box D PMS-JW-007D n Control Room rator Workstation A N/A rator Workstation B N/A ervisor Workstation N/A tch Station (Including Switches) N/A PS MCR Display Unit PMS-JY-001B PS MCR Display Unit PMS-JY-001C R Load Shed Panel 1 VES-EP-01 R Load Shed Panel 2 VES-EP-02 ive Valves tainment Isolation - Air Out olenoid Valve CAS-PL-V014-S imit Switch CAS-PL-V014-L tainment Isolation - Inlet imit Switch CCS-PL-V200-L otor Operator CCS-PL-V200-M tainment Isolation - Outlet imit Switch CCS-PL-V207-L otor Operator CCS-PL-V207-M tainment Isolation - Outlet imit Switch CCS-PL-V208-L otor Operator CCS-PL-V208-M 3I-22 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number S Purification Stop Valve imit Switch CVS-PL-V001-L otor Operator CVS-PL-V001-M S Purification Stop Valve imit Switch CVS-PL-V002-L otor Operator CVS-PL-V002-M S Letdown Stop Valve imit Switch CVS-PL-V003-L otor Operator CVS-PL-V003-M S Letdown IRC Isolation imit Switch CVS-PL-V045-L olenoid Valve CVS-PL-V045-S1 down Flow ORC Isolation imit Switch CVS-PL-V047-L olenoid Valve CVS-PL-V047-S1 iliary PZR Spray Isolation imit Switch CVS-PL-V084-L olenoid Valve CVS-PL-V084-S eup Line Containment Isolation imit Switch CVS-PL-V090-L otor Operator CVS-PL-V090-M eup Line Containment Isolation imit Switch CVS-PL-V091-L otor Operator CVS-PL-V091-M 3I-23 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number rogen Addition Containment Isolation imit Switch CVS-PL-V092-L olenoid Valve CVS-PL-V092-S mineralizer Water System Isolation imit Switch CVS-PL-V136A-L olenoid Valve CVS-PL-V136A-S mineralized Water System Isolation imit Switch CVS-PL-V136B-L olenoid Valve CVS-PL-V136B-S CWST Isolation Valve imit Switch PCS-PL-V001A-L olenoid Valve PCS-PL-V001A-S1 CWST Isolation Valve imit Switch PCS-PL-V001B-L olenoid Valve PCS-PL-V001B-S1 CWST Isolation Valve imit Switch PCS-PL-V001C-L otor Operator PCS-PL-V001C-M CWST Isolation Valve imit Switch PCS-PL-V002A-L otor Operator PCS-PL-V002A-M CWST Isolation Valve imit Switch PCS-PL-V002B-L otor Operator PCS-PL-V002B-M 3I-24 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number CWST Isolation Valve imit Switch PCS-PL-V002C-L otor Operator PCS-PL-V002C-M tainment Isolation - Air Sample Line imit Switch PSS-PL-V008-L olenoid Operator PSS-PL-V008-S tainment Isolation - Liquid Sample Line imit Switch PSS-PL-V010A-L olenoid Operator PSS-PL-V010A-S tainment Isolation - Liquid Sample Line imit Switch PSS-PL-V010B-L olenoid Operator PSS-PL-V010B-S tainment Isolation - Liquid Sample Line imit Switch PSS-PL-V011-L olenoid Valve PSS-PL-V011-S tainment Isolation - Sample Return Line imit Switch PSS-PL-V023-L olenoid Valve PSS-PL-V023-S tainment Isolation - Air Sample Line imit Switch PSS-PL-V046-L olenoid Valve PSS-PL-V046-S e Makeup Tank A Discharge Isolation imit Switch PXS-PL-V014A-L olenoid Valve PXS-PL-V014A-S1 3I-25 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number e Makeup Tank B Discharge Isolation imit Switch PXS-PL-V014B-L olenoid Valve PXS-PL-V014B-S1 e Makeup Tank A Discharge Isolation imit Switch PXS-PL-V015A-L olenoid Valve PXS-PL-V015A-S1 e Makeup Tank B Discharge Isolation imit Switch PXS-PL-V015B-L olenoid Valve PXS-PL-V015B-S1 ogen Supply Outside Containment Isolation imit Switch PXS-PL-V042-L olenoid Valve PXS-PL-V042-S HR HX Discharge Isolation imit Switch PXS-PL-V108A-L olenoid Valve PXS-PL-V108A-S1 HR HX Discharge Isolation imit Switch PXS-PL-V108B-L olenoid Valve PXS-PL-V108B-S1 irc Sump A Isolation imit Switch PXS-PL-V118A-L quib Operator PXS-PL-V118A-T irc Sump B Isolation imit Switch PXS-PL-V118B-L quib Operator PXS-PL-V118B-T 3I-26 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number irc Sump A imit Switch PXS-PL-V120A-L quib Operator PXS-PL-V120A-T irc Sump B imit Switch PXS-PL-V120B-L quib Operator PXS-PL-V120B-T ST Injection A imit Switch PXS-PL-V123A-L quib Operator PXS-PL-V123A-T ST Injection B imit Switch PXS-PL-V123B-L quib Operator PXS-PL-V123B-T ST Injection A imit Switch PXS-PL-V125A-L quib Operator PXS-PL-V125A-T ST Injection B imit Switch PXS-PL-V125B-L quib Operator PXS-PL-V125B-T ST Gutter Drain Isolation A imit Switch PXS-PL-V130A-L olenoid Valve PXS-PL-V130A-S1 ST Gutter Drain Isolation B imit Switch PXS-PL-V130B-L olenoid Valve PXS-PL-V130B-S1 3I-27 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number t Stage ADS imit Switch RCS-PL-V001A-L otor Operator RCS-PL-V001A-M t Stage ADS imit Switch RCS-PL-V001B-L otor Operator RCS-PL-V001B-M ond Stage ADS imit Switch RCS-PL-V002A-L otor Operator RCS-PL-V002A-M ond Stage ADS imit Switch RCS-PL-V002B-L otor Operator RCS-PL-V002B-M d Stage ADS imit Switch RCS-PL-V003A-L otor Operator RCS-PL-V003A-M d Stage ADS imit Switch RCS-PL-V003B-L otor Operator RCS-PL-V003B-M rth Stage ADS imit Switch RCS-PL-V004A-L quib Operator RCS-PL-V004A-T rth Stage ADS imit Switch RCS-PL-V004B-L quib Operator RCS-PL-V004B-T 3I-28 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number rth Stage ADS imit Switch RCS-PL-V004C-L quib Operator RCS-PL-V004C-T rth Stage ADS imit Switch RCS-PL-V004D-L quib Operator RCS-PL-V004D-T t Stage ADS Isolation imit Switch RCS-PL-V011A-L otor Operator RCS-PL-V011A-M t Stage ADS Isolation imit Switch RCS-PL-V011B-L otor Operator RCS-PL-V011B-M ond Stage ADS Isolation imit Switch RCS-PL-V012A-L otor Operator RCS-PL-V012A-M ond Stage ADS Isolation imit Switch RCS-PL-V012B-L otor Operator RCS-PL-V012B-M d Stage ADS Isolation imit Switch RCS-PL-V013A-L otor Operator RCS-PL-V013A-M d Stage ADS Isolation imit Switch RCS-PL-V013B-L otor Operator RCS-PL-V013B-M 3I-29 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number ctor Vessel Head Vent imit Switch RCS-PL-V150A-L olenoid Operator RCS-PV-V150A-S ctor Vessel Head Vent imit Switch RCS-PL-V150B-L olenoid Operator RCS-PL-V150B-S ctor Vessel Head Vent imit Switch RCS-PL-V150C-L olenoid Operator RCS-PL-V150C-S ctor Vessel Head Vent imit Switch RCS-PL-V150D-L olenoid Operator RCS-PL-V150D-S S Inner Suction Isolation imit Switch RNS-PL-V001A-L otor Operator RNS-PL-V001A-M S Inner Suction Isolation imit Switch RNS-PL-V001B-L otor Operator RNS-PL-V001B-M S Outer Suction Isolation imit Switch RNS-PL-V002A-L otor Operator RNS-PL-V002A-M S Outer Suction Isolation imit Switch RNS-PL-V002B-L otor Operator RNS-PL-V002B-M 3I-30 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number R Control/Isolation Valve imit Switch RNS-PL-V011-L otor Operator RNS-PL-V011-M R Pump Suction Header Isolation imit Switch RNS-PL-V022-L otor Operator RNS-PL-V022-M ST Suction Line Isolation imit Switch RNS-PL-V023-L otor Operator RNS-PL-V023-M S - CVS Containment Isolation imit Switch RNS-PL-V061-L ir Operator RNS-PL-V061-S S - MCR Isolation imit Switch SDS-PL-V001-L otor Operator SDS-PL-V001-M S - MCR Isolation imit Switch SDS-PL-V002-L otor Operator SDS-PL-V002-M tainment Isolation imit Switch SFS-PL-V034-L otor Operator SFS-PL-V034-M tainment Isolation imit Switch SFS-PL-V035-L otor Operator SFS- PL-V035-M tainment Isolation imit Switch SFS-PL-V038-L otor Operator SFS-PL-V038-M 3I-31 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number RV Block Valve imit Switch SGS-PL-V027A-L otor Operator SGS-PL-V027A-M RV Block Valve imit Switch SGS-PL-V027B-L otor Operator SGS-PL-V027B-M am Line Condensate Drain Isolation imit Switch SGS-PL-V036A-L olenoid Valve SGS-PL-V036A-S am Line Condensate Isolation imit Switch SGS-PL-V036B-L olenoid Valve SGS-PL-V036B-S n Steam Line Isolation imit Switch SGS-PL-V040A-L olenoid Valve SGS-PL-V040A-S1 olenoid Valve SGS-PL-V040A-S2 olenoid Valve SGS-PL-V040A-S3 olenoid Valve SGS-PL-V040A-S4 n Steam Line Isolation imit Switch SGS-PL-V040B-L olenoid Valve SGS-PL-V040B-S1 olenoid Valve SGS-PL-V040B-S2 olenoid Valve SGS-PL-V040B-S3 olenoid Valve SGS-PL-V040B-S4 3I-32 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number n Feedwater Isolation imit Switch SGS-PL-V057A-L olenoid Valve SGS-PL-V057A-S1 olenoid Valve SGS-PL-V057A-S2 olenoid Valve SGS-PL-V057A-S3 olenoid Valve SGS-PL-V057A-S4 n Feedwater Isolation imit Switch SGS-PL-V057B-L olenoid Valve SGS-PL-V057B-S1 olenoid Valve SGS-PL-V057B-S2 olenoid Valve SGS-PL-V057B-S3 olenoid Valve SGS-PL-V057B-S4 tup Feedwater Isolation imit Switch SGS-PL-V067A-L otor Operator SGS-PL-V067A-M tup Feedwater Isolation imit Switch SGS-PL-V067B-L otor Operator SGS-PL-V067B-M Blowdown Isolation imit Switch SGS-PL-V074A-L olenoid Valve SGS-PL-V074A-S Blowdown Isolation imit Switch SGS-PL-V074B-L olenoid Valve SGS-PL-V074B-S 3I-33 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number Series Blowdown Isolation imit Switch SGS-PL-V075A-L olenoid Valve SGS-PL-V075A-S Series Blowdown Isolation imit Switch SGS-PL-V075B-L olenoid Valve SGS-PL-V075B-S am Line Condensate Drain Isolation Solenoid Valve SGS-PL-V086A-S am Line Condensate Drain Isolation Solenoid Valve SGS-PL-V086B-S er Operated Relief Valve imit Switch SGS-PL-V233A-L olenoid Valve SGS-PL-V233A-S er Operated Relief Valve imit Switch SGS-PL-V233B-L olenoid Valve SGS-PL-V233B-S V Bypass Isolation Valve imit Switch SGS- PL-V240A-L olenoid Valve SGS-PL-V240A-S1 olenoid Valve SGS-PL-V240A-S2 V Bypass Isolation Valve imit Switch SGS-PL-V240B-L olenoid Valve SGS-PL-V240B-S1 olenoid Valve SGS-PL-V240B-S2 3I-34 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number n Feedwater Control Valve imit Switch SGS-PL-V250A-L olenoid Valve SGS-PL-V250A-S n Feedwater Control Valve imit Switch SGS-PL-V250B-L olenoid Valve SGS-PL-V250B-S tup Feedwater Control Valve imit Switch SGS-PL-V255A-L olenoid Valve SGS-PL-V255A-S tup Feedwater Control Valve imit Switch SGS-PL-V255B-L olenoid Valve SGS-PL-V255B-S R Isolation Valve imit Switch VBS-PL-V186-L otor Operator VBS-PL-V186-M R Isolation Valve imit Switch VBS-PL-V187-L otor Operator VBS-PL-V187-M R Isolation Valve imit Switch VBS-PL-V188-L otor Operator VBS-PL-V188-M R Isolation Valve imit Switch VBS-PL-V189-L otor Operator VBS-PL-V189-M 3I-35 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number R Isolation Valve imit Switch VBS-PL-V190-L otor Operator VBS-PL-V190-M R Isolation Valve imit Switch VBS-PL-V191-L otor Operator VBS-PL-V191-M uation Valve A imit Switch VES-PL-V005A-L olenoid Operator VES-PL-V005A-S uation Valve B imit Switch VES-PL-V005B-L olenoid Operator VES-PL-V005B-S ef Isolation Valve A imit Switch VES-PL-V022A-L olenoid Valve VES-PL-V022A-S ef Isolation Valve B imit Switch VES-PL-V022B-L olenoid Valve VES-PL-V022B-S tainment Purge Inlet Isolation imit Switch VFS-PL-V003-L olenoid Valve VFS-PL-V003-S1 tainment Purge Inlet Isolation imit Switch VFS-PL-V004-L olenoid Valve VFS-PL-V004-S1 3I-36 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number tainment Purge Discharge Isolation imit Switch VFS-PL-V009-L olenoid Valve VFS-PL-V009-S1 tainment Purge Discharge Isolation imit Switch VFS-PL-V010-L olenoid Valve VFS-PL-V010-S1 uum Relief Containment Isolation Valve A - ORC imit Switch VFS-PL-V800A-L otor Operator VFS-PL-V800A-M uum Relief Containment Isolation Valve B - ORC imit Switch VFS-PL-V800B-L otor Operator VFS-PL-V800B-M Cooler Supply Isolation imit Switch VWS-PL-V058-L olenoid Valve VWS-PL-V058-S Cooler Return Isolation imit Switch VWS-PL-V082-L olenoid Valve VWS-PL-V082-S Cooler Return Isolation imit Switch VWS-PL-V086-L olenoid Valve VWS-PL-V086-S p Containment Isolation IRC imit Switch WLS-PL-V055-L olenoid Valve WLS-PL-V055-S1 3I-37 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number p Containment Isolation ORC imit Switch WLS-PL-V057-L olenoid Valve WLS-PL-V057-S1 DT Gas Containment Isolation imit Switch WLS-PL-V067-L olenoid Valve WLS-PL-V067-S DT Gas Containment Isolation imit Switch WLS-PL-V068-L olenoid Valve WLS-PL-V068-S Leg 1 Sample Isolation imit Switch PSS-PL-V001A-L Leg 2 Sample Isolation imit Switch PSS-PL-V001B-L e Makeup Tank A CL Inlet Isolation imit Switch PXS-PL-V002A-L otor Operator PXS-PL-V002A-M e Makeup Tank B CL Inlet Isolation imit Switch PXS-PL-V002B-L otor Operator PXS-PL-V002B-M HR HX Inlet Isolation imit Switch PXS-PL-V101-L otor Operator PXS-PL-V101-M irc Sump A Isolation imit Switch PXS-PL-V117A-L otor Operator PXS-PL-V117A-M 3I-38 Revision 1
Electro-mechanical Equipment AP1000 Description Tag Number irc Sump B Isolation imit Switch PXS-PL-V117B-L otor Operator PXS-PL-V117B-M rth Stage ADS Isolation imit Switch RCS-PL-V014A-L otor Operator RCS-PL-V014A-M rth Stage ADS Isolation imit Switch RCS-PL-V014B-L otor Operator RCS-PL-V014B-M rth Stage ADS Isolation imit Switch RCS-PL-V014C-L otor Operator RCS-PL-V014C-M rth Stage ADS Isolation imit Switch RCS-PL-V014D-L otor Operator RCS-PL-V014D-M 3I-39 Revision 1
and Mechanical Equipment Not High Frequency Sensitive AP1000 Description Tag Number Comment istance Temperature Detectors HR HX Outlet Temperature RCS-JE-TE161 1 S Cold Leg 1A Narrow Range Temperature RCS-JE-TE121A 1 S Cold Leg 1A Narrow Range Temperature RCS-JE-TE121D 1 S Cold Leg 1B Narrow Range Temperature RCS-JE-TE121B 1 S Cold Leg 1B Narrow Range Temperature RCS-JE-TE121C 1 S Cold Leg 2A Narrow Range Temperature RCS-JE-TE122B 1 S Cold Leg 2A Narrow Range Temperature RCS-JE-TE122C 1 S Cold Leg 2B Narrow Range Temperature RCS-JE-TE122A 1 S Cold Leg 2B Narrow Range Temperature RCS-JE-TE122D 1 S Hot Leg 1 Narrow Range Temperature RCS-JE-TE131A 1 S Hot Leg 1 Narrow Range Temperature RCS-JE-TE131C 1 S Hot Leg 1 Narrow Range Temperature RCS-JE-TE132A 1 S Hot Leg 1 Narrow Range Temperature RCS-JE-TE132C 1 S Hot Leg 1 Narrow Range Temperature RCS-JE-TE133C 1 S Hot Leg 1 Narrow Range Temperature RCS-JE-TE133A 1 S Hot Leg 2 Narrow Range Temperature RCS-JE-TE131B 1 S Hot Leg 2 Narrow Range Temperature RCS-JE-TE131D 1 S Hot Leg 2 Narrow Range Temperature RCS-JE-TE132B 1 S Hot Leg 2 Narrow Range Temperature RCS-JE-TE132D 1 S Hot Leg 2 Narrow Range Temperature RCS-JE-TE133B 1 S Hot Leg 2 Narrow Range Temperature RCS-JE-TE133D 1 S Cold Leg 1A Dual Range Temperature RCS-JE-TE125A 1 S Cold Leg 1B Dual Range Temperature RCS-JE-TE125C 1 S Cold Leg 2A Dual Range Temperature RCS-JE-TE125B 1 3I-40 Revision 1
AP1000 Description Tag Number Comment S Cold Leg 2B Dual Range Temperature RCS-JE-TE125D 1 S Hot Leg 1 Wide Range Temperature RCS-JE-TE135A 1 S Hot Leg 2 Wide Range Temperature RCS-JE-TE135B 1 Reference Leg Level Temperature RCS-JE-TE193A 1 Reference Leg Level Temperature RCS-JE-TE193B 1 Reference Leg Level Temperature RCS-JE-TE193C 1 Reference Leg Level Temperature RCS-JE-TE193D 1 rmocouples 1 re Thermocouples IIS-JE-TE001-TE042 1 P 1A Bearing Water Temperature RCS-JE-TE211A 1 P 1A Bearing Water Temperature RCS-JE-TE211B 1 P 1A Bearing Water Temperature RCS-JE-TE211C 1 P 1A Bearing Water Temperature RCS-JE-TE211D 1 P 1B Bearing Water Temperature RCS-JE-TE212A 1 P 1B Bearing Water Temperature RCS-JE-TE212B 1 P 1B Bearing Water Temperature RCS-JE-TE212C 1 P 1B Bearing Water Temperature RCS-JE-TE212D 1 P 2A Bearing Water Temperature RCS-JE-TE213A 1 P 2A Bearing Water Temperature RCS-JE-TE213B 1 P 2A Bearing Water Temperature RCS-JE-TE213C 1 P 2A Bearing Water Temperature RCS-JE-TE213D 1 P 2B Bearing Water Temperature RCS-JE-TE214A 1 P 2B Bearing Water Temperature RCS-JE-TE214B 1 P 2B Bearing Water Temperature RCS-JE-TE214C 1 P 2B Bearing Water Temperature RCS-JE-TE214D 1 3I-41 Revision 1
AP1000 Description Tag Number Comment etrations etrations (Mechanical) 1 etrations (Electrical) 1 ive Valves tainment Isolation - Air Out CAS-PL-V014 2 tainment Isolation - Air In CAS-PL-V015 2 tainment Isolation - Inlet CCS-PL-V200 2 vice Air Supply Inside Containment Isolation CAS-PL-V205 2 tainment Isolation - Inlet CCS-PL-V201 2 tainment Isolation - Outlet CCS-PL-V207 2 tainment Isolation - Outlet CCS-PL-V208 2 S Containment Isolation Relief CCS-PL-V220 2 S IRC Relief Valve CCS-PL-V270 2 S IRC Relief Valve CCS-PL-V271 2 S Purification Stop Valve CVS-PL-V001 2 S Purification Stop Valve CVS-PL-V002 2 S Letdown Stop Valve CVS-PL-V003 2 mineralizer Flush Line Relief Valve CVS-PL-V042 2 S Letdown IRC Isolation CVS-PL-V045 2 down Flow ORC Isolation CVS-PL-V047 2 S Purification Check Valve CVS-PL-V080 2 S Purification Stop Valve CVS-PL-V081 2 S Purification Check Valve CVS-PL-V082 2 iliary PZR Spray Isolation CVS-PL-V084 2 iliary PZR Spray Isolation CVS-PL-V085 2 eup Line Containment Isolation CVS-PL-V090 2 eup Line Containment Isolation CVS-PL-V091 2 3I-42 Revision 1
AP1000 Description Tag Number Comment rogen Addition Containment Isolation CVS-PL-V092 2 rogen Addition Containment Isolation CVS-PL-V094 2 rogen Addition Containment Isolation CVS-PL-V092 2 rogen Addition Containment Isolation CVS-PL-V094 2 eup Containment Isolation CVS-PL-V100 2 mineralizer Water System Isolation CVS-PL-V136A 2 mineralized Water System Isolation CVS-PL-V136B 2 min Water Supply Containment Isolation - Inside DWS-PL-V245 2 l Transfer Tube Gate Valve FHS-PL-V001 2 Water Containment Supply Isolation - Inside FPS-PL-V052 2 CWST Isolation Valve PCS-PL-V001A 2 CWST Isolation Valve PCS-PL-V001B 2 CWST Isolation Valve PCS-PL-V001C 2 CWST Isolation Valve PCS-PL-V002A 2 CWST Isolation Valve PCS-PL-V002B 2 CWST Isolation Valve PCS-PL-V002C 2 CWST Fire Protection Isolation PCS-PL-V005 2 CWST Emergency Spent Fuel Pool Makeup Isolation PCS-PL-V009 2 er Bucket Makeup Line Drain Valve PCS-PL-V015 2 er Bucket Makeup Line Isolation Valve PCS-PL-V020 2 S Recirculation Isolation PCS-PL-V023 2 CWST Long-Term Makeup Check Valve PCS-PL-V039 2 CWST Long Term Makeup Isolation Drain Valve PCS-PL-V042 2 CWST Long Term Makeup Isolation Valve PCS-PL-V044 2 ergency Makeup to the Spent Fuel Pool Isolation Valve PCS-PL-V045 2 3I-43 Revision 1
AP1000 Description Tag Number Comment CWST Recirculation Return Isolation Valve PCS-PL-V046 2 ergency Makeup to the Spent Fuel Pool Drain Isolation Valve PCS-PL-V049 2 nt Fuel Pool Long Term Makeup Isolation Valve PCS-PL-V050 2 nt Fuel Pool Emergency Makeup Lower Isolation Valve PCS-PL-V051 2 tainment Isolation - Air Sample Line PSS-PL-V008 2 tainment Isolation - Liquid Sample Line PSS-PL-V010A 2 tainment Isolation - Liquid Sample Line PSS-PL-V010B 2 tainment Isolation - Liquid Sample Line PSS-PL-V011 2 tainment Isolation - Sample Return Line PSS-PL-V023 2 tainment Isolation Sample Return PSS-PL-V024 2 tainment Isolation - Air Sample Line PSS-PL-V046 2 S MCR Isolation PWS-PL-V418 2 S MCR Isolation PWS-PL-V420 2 S MCR Vacuum Relief PWS-PL-V498 2 e Makeup Tank A Discharge Isolation PXS-PL-V014A 2 e Makeup Tank B Discharge Isolation PXS-PL-V014B 2 e Makeup Tank A Discharge Isolation PXS-PL-V015A 2 e Makeup Tank B Discharge Isolation PXS-PL-V015B 2 e Makeup Tank A Discharge PXS-PL-V016A 2 e Makeup Tank B Discharge PXS-PL-V016B 2 e Makeup Tank A Discharge PXS-PL-V017A 2 e Makeup Tank B Discharge PXS-PL-V017B 2 umulator A Pressure Relief PXS-PL-V022A 2 umulator B Pressure Relief PXS-PL-V022B 2 umulator A Discharge PXS-PL-V028A 2 3I-44 Revision 1
AP1000 Description Tag Number Comment umulator B Discharge PXS-PL-V028B 2 umulator A Discharge PXS-PL-V029A 2 umulator B Discharge PXS-PL-V029B 2 ogen Supply Outside Containment Isolation PXS-PL-V042 2 Nitrogen Supply Inside Containment Isolation PXS-PL-V043 2 HR HX Discharge Isolation PXS-PL-V108A 2 HR HX Discharge Isolation PXS-PL-V108B 2 irc Sump A Isolation PXS-PL-V118A 2 irc Sump B Isolation PXS-PL-V118B 2 irc Sump A Isolation PXS-PL-V119A 2 irc Sump B Isolation PXS-PL-V119B 2 irc Sump A Isolation PXS-PL-V120A 2 irc Sump B Isolation PXS- PL-V120B 2 ST Injection A PXS-PL-V122A 2 ST Injection B PXS-PL-V122B 2 ST Injection A PXS-PL-V123A 2 ST Injection B PXS-PL-V123B 2 ST Injection A PXS-PL-V124A 2 ST Injection B PXS-PL-V124B 2 ST Injection A PXS-PL-V125A 2 ST Injection B PXS-PL-V125B 2 ST Gutter Drain Isolation A PXS-PL-V130A 2 ST Gutter Drain Isolation B PXS-PL-V130B 2 t Stage ADS RCS-PL-V001A 2 t Stage ADS RCS-PL-V001B 2 3I-45 Revision 1
AP1000 Description Tag Number Comment ond Stage ADS RCS-PL-V002A 2 ond Stage ADS RCS-PL-V002B 2 d Stage ADS RCS-PL-V003A 2 d Stage ADS RCS-PL-V003B 2 rth Stage ADS RCS-PL-V004A 2 rth Stage ADS RCS-PL-V004B 2 rth Stage ADS RCS-PL-V004C 2 rth Stage ADS RCS-PL-V004D 2 Safety Valve RCS-PL-V005A 2 Safety Valve RCS-PL-V005B 2 S Discharge Header A Relief RCS-PL-V010A 2 S Discharge Header B Relief RCS-PL-V010B 2 t Stage ADS Isolation RCS-PL-V011A 2 t Stage ADS Isolation RCS-PL-V011B 2 ond Stage ADS Isolation RCS-PL-V012A 2 ond Stage ADS Isolation RCS-PL-V012B 2 d Stage ADS Isolation RCS-PL-V013A 2 d Stage ADS Isolation RCS-PL-V013B 2 ctor Vessel Head Vent RCS-PL-V150A 2 ctor Vessel Head Vent RCS-PL-V150B 2 ctor Vessel Head Vent RCS-PL-V150C 2 ctor Vessel Head Vent RCS-PL-V150D 2 S Inner Suction Isolation RNS-PL-V001A 2 S Inner Suction Isolation RNS-PL-V001B 2 S Outer Suction Isolation RNS-PL-V002A 2 3I-46 Revision 1
AP1000 Description Tag Number Comment S Outer Suction Isolation RNS-PL-V002B 2 S Thermal Relief RNS-PL-V003A 2 S Thermal Relief RNS-PL-V003B 2 R Control/Isolation Valve RNS-PL-V011 2 S Discharge Containment Isolation Valve Test Connection RNS-PL-V012 2 S Discharge Containment Isolation RNS-PL-V013 2 S Discharge RCP B Isolation RNS-PL-V015A 2 S Discharge RCP B Isolation RNS-PL-V015B 2 S Discharge RCP B Isolation RNS-PL-V017A 2 S Discharge RCP B Isolation RNS-PL-V017B 2 S Hot Leg Suction Relief RNS-PL-V021 2 R Pump Suction Header Isolation RNS-PL-V022 2 ST Suction Line Isolation RNS-PL-V023 2 S Pump Discharge Relief RNS-PL-V045 2 S - CVS Containment Isolation RNS-PL-V061 2 tainment Isolation SFS-PL-V034 2 tainment Isolation SFS-PL-V035 2 Discharge Containment Isolation SFS-PL-V037 2 tainment Isolation SFS-PL-V038 2 Cask Loading Pit to SFS Pump SFS-PL-V042 2 Pump to Cask Loading Pit SFS-PL-V045 2 k Loading Pit to WLS SFS-PL-V049 2 nt Fuel Pool to Cask Washdown Pit Isolation SFS-PL-V066 2 Containment Isolation Relief SFS-PL-V067 2 k Washdown Pit Drain Isolation SFS-PL-V068 2 3I-47 Revision 1
AP1000 Description Tag Number Comment ueling Cavity to SG Compartment SFS-PL-V071 2 ueling Cavity to SG Compartment SFS-PL-V072 2 RV Block Valve SGS-PL-V027A 2 RV Block Valve SGS-PL-V027B 2 am Safety Valve SG01 SGS-PL-V030A 2 am Safety Valve SG02 SGS-PL-V030B 2 am Safety Valve SG01 SGS-PL-V031A 2 am Safety Valve SG02 SGS-PL-V031B 2 am Safety Valve SG01 SGS-PL-V032A 2 am Safety Valve SG02 SGS-PL-V032B 2 am Safety Valve SG01 SGS-PL-V033A 2 am Safety Valve SG02 SGS-PL-V033B 2 am Safety Valve SG01 SGS-PL-V034A 2 am Safety Valve SG02 SGS-PL-V034B 2 am Safety Valve SG01 SGS-PL-V035A 2 am Safety Valve SG02 Steam Line Condensate SGS-PL-V035B 2 in Isolation SGS-PL-V036A 2 am Line Condensate Isolation SGS-PL-V036B 2 n Steam Line Isolation SGS-PL-V040A 2 n Steam Line Isolation SGS-PL-V040B 2 n Feedwater Isolation SGS-PL-V057A 2 n Feedwater Isolation SGS-PL-V057B 2 tup Feedwater Isolation SGS-PL-V067A 2 tup Feedwater Isolation SGS-PL-V067B 2 Blowdown Isolation SGS-PL-V074A 2 3I-48 Revision 1
AP1000 Description Tag Number Comment Blowdown Isolation SGS-PL-V074B 2 Series Blowdown Isolation SGS-PL-V075A 2 Series Blowdown Isolation SGS-PL-V075B 2 am Line Condensate Drain Isolation SGS-PL-V086A 2 am Line Condensate Drain Isolation SGS-PL-V086B 2 er-Operated Relief Valve SGS-PL-V233A 2 er-Operated Relief Valve SGS-PL-V233B 2 V Bypass Isolation Valve SGS-PL-V240A 2 V Bypass Isolation Valve SGSPL-V240B 2 n Feedwater Control Valve SGS-PL-V250A 2 n Feedwater Control Valve SGS-PL-V250B 2 tup Feedwater Control Valve SGS-PL-V255A 2 tup Feedwater Control Valve SGS-PL-V255B 2 R Isolation Valve VBS-PL-V186 2 R Isolation Valve VBS-PL-V187 2 R Isolation Valve VBS-PL-V188 2 R Isolation Valve VBS-PL-V189 2 R Isolation Valve VBS-PL-V190 2 R Isolation Valve VBS-PL-V191 2 Delivery Isolation Valve VES-PL-V001 2 ssure Regulator Valve A VES-PL-V002A 2 ssure Regulator Valve B VES-PL-V002B 2 uation Valve A VES-PL-V005A 2 uation Valve B VES-PL-V005B 2 porary Instrument Isolation Valve A VES-PL-V018 2 porary Instrument Isolation Valve B VES-PL-V019 2 ef Isolation Valve A VES-PL-V022A 2 ef Isolation Valve B VES-PL-V022B 2 3I-49 Revision 1
AP1000 Description Tag Number Comment Tank Relief A VES-PL-V040A 2 Tank Relief B VES-PL-V040B 2 Tank Relief A VES-PL-V041A 2 Tank Relief B VES-PL-V041B 2 n Air Flow Path Isolation Valve VES-PL-V044 2 tainment Purge Inlet Isolation VFS-PL-V003 2 tainment Purge Inlet Isolation VFS-PL-V004 2 tainment Purge Discharge Isolation VFS-PL-V009 2 tainment Purge Discharge Isolation VFS-PL-V010 2 uum Relief Containment Isolation Valve A - ORC VFS-PL-V800A 2 uum Relief Containment Isolation Valve B - ORC VFS-PL-V800B 2 uum Relief Containment Isolation Check Valve A - IRC VFS-PL-V803A 2 uum Relief Containment Isolation Check Valve B - IRC VFS-PL-V803B 2 Cooler Supply Isolation VWS-PL-V058 2 Cooler Supply Isolation VWS-PL-V062 2 S Containment Isolation Relief VWS-PL-V080 2 Cooler Return Isolation VWS-PL-V082 2 Cooler Return Isolation VWS-PL-V086 2 p Containment Isolation IRC WLS-PL-V055 2 p Containment Isolation ORC WLS-PL-V057 2 S Containment Isolation Relief WLS-PL-V058 2 DT Gas Containment Isolation WLS-PL-V067 2 DT Gas Containment Isolation WLS-PL-V068 2 S To Sump WLS-PL-V071 A 2 A To Sump WLS-PL-V071 B 2 3I-50 Revision 1
AP1000 Description Tag Number Comment B To Sump WLS-PL-V071 C 2 S To Sump WLS-PL-V072 A 2 A To Sump WLS-PL-V072 B 2 B To Sump WLS-PL-V072 C 2 cellaneous active Valves tainment Penetration Test Connection Isolation CAS-PL-V027 2 vice Air Supply Outside Containment Isolation CAS-PL-V204 2 tainment Penetration Test Connection Isolation CAS-PL-V219 2 tainment Isolation Valve Test Connection - Outlet Line CCS-PL-V209 2 S Supply Containment Isolation - IRC CCS-PL-V214 2 S Supply Containment Isolation Valve Test Connection - IRC CCS-PL-V215 2 tainment Leak Test Outlet Line - IRC CCS-PL-V216 2 tainment Isolation Valve V207 Body Test Connection Valve CCS-PL-V217 2 tainment Isolation Valve Test Connection - Inlet Line CCS-PL-V257 2 in Flush IRC Isolation CVS-PL-V040 2 in Flush ORC Isolation CVS-PL-V041 2 down PZR Instrument Root CVS-PL-V046 2 Mkup Containment Isolation Thermal Relief Valve CVS-PL-V065 2 rogen Add Cont Isolation Test Connection CVS-PL-V095 2 rogen Addition Containment Isolation Test Connection CVS-PL-V096 2 min Water Supply Containment Isolation - Outside DWS-PL-V244 2 tainment Penetration Test Connection Isolation DWS-PL-V248 2 Water Containment Test Connection Isolation FPS-PL-V049 2 Water Containment Supply Isolation FPS-PL-V050 2 3I-51 Revision 1
AP1000 Description Tag Number Comment Water Containment Test Connection Isolation FPS-PL-V051 2 w Transmitter FT001 Root Valve PCS-PL-V010A 2 w Transmitter FT001 Root Valve PCS-PL-V010B 2 w Transmitter FT002 Root Valve PCS-PL-V011A 2 w Transmitter FT001 Root Valve PCS-PL-V011B 2 w Transmitter FT003 Root Valve PCS-PL-V012A 2 w Transmitter FT003 Root Valve PCS-PL-V012B 2 w Transmitter FT004 Root Valve PCS-PL-V013A 2 w Transmitter FT004 Root Valve PCS-PL-V013B 2 CWST Drain Isolation Valve PCS-PL-V016 2 CWST Isolation Valve Leakage Detection Drain PCS-PL-V029 2 CWST Isolation Valve Leakage Detection Crossconn PCS-PL-V030 2 CWST Level Instrument Root Valve PCS-PL-V031A 2 CWST Level Instrument Root Valve PCS-PL-V031B 2 irculation Pump Suction from Long Term Makeup Isolation Valve PCS-PL-V033 2 nt Fuel Pool Emergency Makeup Isolation PLS-PL-V052 2 Leg 1 Sample Isolation PSS-PL-V001A 2 Leg 2 Sample Isolation PSS-PL-V001B 2 ssurizer Sample Isolation PSS-PL-V003 2 S Accumulator Sample Isolation PSS-PL-V004A 2 Accumulator Sample Isolation PSS-PL-V004B 2 S CMT A Sample Isolation PSS-PL-V005A 2 S CMT B Sample Isolation PSS-PL-V005B 2 S CMT A Sample Isolation PSS-PL-V005C 2 3I-52 Revision 1
AP1000 Description Tag Number Comment S CMT B Sample Isolation PSS-PL-V005D 2 id Sample Check Valve PSS-PL-V012A 2 id Sample Check Valve PSS-PL-V012B 2 tainment Testing Boundary Isolation Valve PSS-PL-V076A 2 tainment Testing Boundary Isolation Valve PSS-PL-V076B 2 tainment Isolation Test Connection Isolation Valve PSS-PL-V082 2 tainment Isolation Test Connection Isolation Valve PSS-PL-V083 2 tainment Isolation Test Connection Isolation Valve PSS-PL-V085 2 tainment Isolation Test Connection Isolation Valve PSS-PL-V086 2 e Makeup Tank A CL Inlet Isolation PXS-PL-V002A 2 e Makeup Tank B CL Inlet Isolation PXS-PL-V002B 2 e Makeup Tank A Upper Sample PXS-PL-V010A 2 e Makeup Tank B Upper Sample PXS-PL-V010B 2 e Makeup Tank A Lower Sample PXS-PL-V011A 2 e Makeup Tank B Lower Sample PXS-PL-V011B 2 e Makeup Tank A Drain PXS-PL-V012A 2 e Makeup Tank B Drain PXS-PL-V012B 2 e Makeup Tank Discharge Manual Isolation PXS-PL-V013A 2 e Makeup Tank B Discharge Manual Isolation PXS-PL-V013B 2 S to CMT Injection Line A Drain PXS-PL-V019A 2 S to CMT Injection Line B Drain PXS-PL-V019B 2 ST Injection Line A Drain PXS-PL-V020A 2 ST Injection Line B Drain PXS-PL-V020B 2 umulator A N2 Vent PXS-PL-V021A 2 umulator B N2 Vent PXS-PL-V021B 2 3I-53 Revision 1
AP1000 Description Tag Number Comment umulator A PZR Transmitter Isolation PXS-PL-V023A 2 umulator B PZR Transmitter Isolation PXS-PL-V023B 2 umulator A PZR Transmitter Isolation PXS-PL-V024A 2 umulator B PZR Transmitter Isolation PXS-PL-V024B 2 umulator A Sample PXS-PL-V025A 2 umulator B Sample PXS-PL-V025B 2 umulator A Drain PXS-PL-V026A 2 umulator B Drain PXS-PL-V026B 2 umulator A Discharge Isolation PXS-PL-V027A 2 umulator B Discharge Isolation PXS-PL-V027B 2 e Makeup Tank A Highpoint Vent PXS-PL-V030A 2 e Makeup Tank B Highpoint Vent PXS-PL-V030B 2 e Makeup Tank A Highpoint Vent PXS-PL-V031A 2 e Makeup Tank B Highpoint Vent PXS-PL-V031B 2 umulator A Check Valve Drain PXS-PL-V033A 2 umulator B Check Valve Drain PXS-PL-V033B 2 umulator N2 Containment Penetration Test Connection PXS-PL-V052 2 T A Wide Level Upper Root PXS-PL-V080A 2 T B Wide Level Upper Root PXS-PL-V080B 2 T A Wide Level Lower Root PXS-PL-V081A 2 T B Wide Level Lower Root PXS-PL-V081B 2 T A Upper Level A Isolation 1 PXS-PL-V082A 2 T B Upper Level A Isolation 1 PXS-PL-V082B 2 T A Upper Level A Isolation 2 PXS-PL-V083A 2 T B Upper Level A Isolation 2 PXS-PL-V083B 2 T A Upper Level A Vent PXS-PL-V084A 2 3I-54 Revision 1
AP1000 Description Tag Number Comment T B Upper Level A Vent PXS-PL-V084B 2 T A Upper Level A Drain PXS-PL-V085A 2 T B Upper Level A Drain PXS-PL-V085B 2 T A Upper Level B Isolation 1 PXS-PL-V086A 2 T B Upper Level B Isolation 1 PXS-PL-V086B 2 T A Upper Level B Isolation 2 PXS-PL-V087A 2 T B Upper Level B Isolation 2 PXS-PL-V087B 2 T A Upper Level B Vent PXS-PL-V088A 2 T B Upper Level B Vent PXS-PL-V088B 2 T A Upper Level B Drain PXS-PL-V089A 2 T B Upper Level B Drain PXS-PL-V089B 2 T A Lower Level A Isolation 1 PXS-PL-V092A 2 T B Lower Level A Isolation 1 PXS-PL-V092B 2 T A Lower Level A Isolation 2 PXS-PL-V093A 2 T B Lower Level A Isolation 2 PXS-PL-V093B 2 T A Lower Level A Vent PXS-PL-V094A 2 T B Lower Level A Vent PXS-PL-V094B 2 T A Lower Level A Drain PXS-PL-V095A 2 T B Lower Level A Drain PXS-PL-V095B 2 T A Lower Level B Isolation 1 PXS-PL-V096A 2 T B Lower Level B Isolation 1 PXS-PL-V096B 2 T A Lower Level B Isolation 2 PXS-PL-V097A 2 T B Lower Level B Isolation 2 PXS-PL-V097B 2 T A Lower Level B Vent PXS-PL-V098A 2 T B Lower Level B Vent PXS-PL-V098B 2 3I-55 Revision 1
AP1000 Description Tag Number Comment T A Lower Level B Drain PXS-PL-V099A 2 T B Lower Level B Drain PXS-PL-V099B 2 HR HX Inlet Isolation PXS-PL-V101 2 HR HX Inlet Head Vent PXS-PL-V102A 2 HR HX Inlet Head Drain PXS-PL-V102B 2 HR HX Outlet Head Vent PXS-PL-V103A 2 HR HX Outlet Head Drain PXS-PL-V103B 2 HR HX Flow Transmitter A Isolation PXS-PL-V104A 2 HR HX Flow Transmitter B Isolation PXS-PL-V104B 2 HR HX Flow Transmitter A Isolation PXS-PL-V105A 2 HR HX Flow Transmitter B Isolation PXS-PL-V105B 2 tainment Recirculation A Highpoint Vent PXS-PL-V106 2 tainment Recirculation A Highpoint Vent PXS-PL-V107 2 HR HX/RCS Return Isolation PXS-PL-V109 2 HR HX Highpoint Vent PXS-PL-V111A 2 HR HX Highpoint Vent PXS-PL-V111B 2 HR HX PZR Transmitter Isolation PXS-PL-V113 2 tainment Recirculation A Drain PXS-PL-V115A 2 tainment Recirculation B Drain PXS-PL-V115B 2 tainment Recirculation A Drain PXS-PL-V116A 2 tainment Recirculation B Drain PXS-PL-V116B 2 irc Sump A Isolation PXS-PL-V117A 2 irc Sump B Isolation PXS-PL-V117B 2 ST Line A Isolation PXS-PL-V121A 2 ST Line B Isolation PXS-PL-V121B 2 ST Injection Check Test PXS-PL-V126A 2 3I-56 Revision 1
AP1000 Description Tag Number Comment ST Injection Check Test PXS-PL-V126B 2 ST Injection Line A Drain PXS-PL-V127 2 ST Injection Check Test PXS-PL-V128A 2 ST Injection Check Test PXS-PL-V128B 2 ST Injection Check Test PXS-PL-V129A 2 ST Injection Check Test PXS-PL-V129B 2 ST Injection Line A Drain PXS-PL-V131A 2 ST Injection Line B Drain PXS-PL-V131B 2 ST Injection Line A Drain PXS-PL-V132A 2 ST Injection Line B Drain PXS-PL-V132B 2 ST Injection Line A Highpoint Vent PXS-PL-V133A 2 ST Injection Line B Highpoint Vent PXS-PL-V133B 2 ST Injection Line A Highpoint Vent PXS-PL-V134A 2 ST Injection Line B Highpoint Vent PXS-PL-V134B 2 ST Injection Line A Highpoint Vent Isolation PXS-PL-V135A 2 ST Injection Line B Highpoint Vent Isolation PXS-PL-V135B 2 S Suction Pump Line Drain PXS-PL-V149 2 ST Level Transmitter A Isolation PXS-PL-V150A 2 ST Level Transmitter B Isolation PXS-PL-V150B 2 ST Level Transmitter C Isolation PXS-PL-V150C 2 ST Level Transmitter D Isolation PXS-PL-V150D 2 ST Level Transmitter A Isolation PXS-PL-V151A 2 ST Level Transmitter B Isolation PXS-PL-V151B 2 ST Level Transmitter C Isolation PXS-PL-V151C 2 ST Level Transmitter D Isolation PXS-PL-V151D 2 umulator A Leak Test PXS-PL-V201A 2 3I-57 Revision 1
AP1000 Description Tag Number Comment umulator B Leak Test PXS-PL-V201B 2 umulator A Leak Test PXS-PL-V202A 2 umulator B Leak Test PXS-PL-V202B 2 S Discharge Leak Test PXS-PL-V205A 2 S Discharge Leak Test PXS-PL-V205B 2 S Discharge Leak Test PXS-PL-V206 2 S Suction Leak Test PXS-PL-V207A 2 S Suction Leak Test PXS-PL-V207B 2 S Suction Leak Test PXS-PL-V208A 2 e Makeup Tank A Fill Isolation PXS-PL-V230A 2 e Makeup Tank B Fill Isolation PXS-PL-V230B 2 e Makeup Tank A Fill Check PXS-PL-V231A 2 e Makeup Tank B Fill Check PXS-PL-V231B 2 umulator A Fill/Drain Isolation PXS-PL-V232A 2 umulator B Fill/Drain Isolation PXS-PL-V232B 2 T A Check Valve Test Valve PXS-PL-V250A 2 T B Check Valve Test Valve PXS-PL-V250B 2 T A Check Valve Test Valve PXS-PL-V251A 2 T B Check Valve Test Valve PXS-PL-V251B 2 T A Check Valve Test Valve PXS-PL-V252A 2 T B Check Valve Test Valve PXS-PL-V252B 2 S Test Valve RCS-PL-V007A 2 S Test Valve RCS-PL-V007B 2 rth Stage ADS Isolation RCS-PL-V014A 2 rth Stage ADS Isolation RCS-PL-V014B 2 rth Stage ADS Isolation RCS-PL-V014C 2 3I-58 Revision 1
AP1000 Description Tag Number Comment rth Stage ADS Isolation RCS-PL-V014D 2 Leg 2 Level Instrument Root RCS-PL-V095 2 Leg 2 Level Instrument Root RCS-PL-V096 2 Leg 1 Level Instrument Root RCS-PL-V097 2 Leg 1 Level Instrument Root RCS-PL-V098 2 Leg 1 Flow Instrument Root RCS-PL-V101A 2 Leg 1 Flow Instrument Root RCS-PL-V101B 2 Leg 1 Flow Instrument Root RCS-PL-V101C 2 Leg 1 Flow Instrument Root RCS-PL-V101D 2 Leg 1 Flow Instrument Root RCS-PL-V101E 2 Leg 1 Flow Instrument Root RCS-PL-V101F 2 Leg 2 Flow Instrument Root RCS-PL-V102A 2 Leg 2 Flow Instrument Root RCS-PL-V102B 2 Leg 2 Flow Instrument Root RCS-PL-V102C 2 Leg 2 Flow Instrument Root RCS-PL-V102D 2 Leg 2 Flow Instrument Root RCS-PL-V102E 2 Leg 2 Flow Instrument Root RCS-PL-V102F 2 HR HX Outlet Line Drain RCS-PL-V103 2 Leg 1 Sample Isolation RCS-PL-V108A 2 Leg 2 Sample Isolation RCS-PL-V108B 2 Spray Valve RCS-PL-V110A 2 Spray Valve RCS-PL-V110B 2 Spray Block Valve RCS-PL-V111A 2 Spray Block Valve RCS-PL-V111B 2 d Leg 1A Bend Instrument Root RCS-PL-V171A 2 3I-59 Revision 1
AP1000 Description Tag Number Comment d Leg 1A Bend Instrument Root RCS-PL-V171B 2 d Leg 1B Bend Instrument Root RCS-PL-V172A 2 d Leg 1B Bend Instrument Root RCS-PL-V172B 2 d Leg 2A Bend Instrument Root RCS-PL-V173A 2 d Leg 2A Bend Instrument Root RCS-PL-V173B 2 d Leg 2B Bend Instrument Root RCS-PL-V174A 2 d Leg 2B Bend Instrument Root RCS-PL-V174B 2 Manual Vent RCS-PL-V204 2 Manual Vent RCS-PL-V205 2 Spray Bypass RCS-PL-V210A 2 Spray Bypass RCS-PL-V210B 2 Level Steam Space Instrument Root RCS-PL-V225A 2 Level Steam Space Instrument Root RCS-PL-V225B 2 Level Steam Space Instrument Root RCS-PL-V225C 2 Level Steam Space Instrument Root RCS-PL-V225D 2 Level Liquid Space Instrument Root RCS-PL-V226A 2 Level Liquid Space Instrument Root RCS-PL-V226B 2 Level Liquid Space Instrument Root RCS-PL-V226C 2 Level Liquid Space Instrument Root RCS-PL-V226D 2 e Range PZR Level Steam Space Instrument Root RCS-PL-V228 2 e Range PZR Level Liquid Space Instrument Root RCS-PL-V229 2 ual Head Vent RCS-PL-V232 2 d Vent Isolation RCS-PL-V233 2 S Valve Discharge Header Drain Isolation RCS-PL-V241 2 P 1A Flush RCS-PL-V260A 2 3I-60 Revision 1
AP1000 Description Tag Number Comment P 1B Flush RCS-PL-V260B 2 P 2A Flush RCS-PL-V260C 2 P 2B Flush RCS-PL-V260D 2 P 1A Drain RCS-PL-V261A 2 P 1B Drain RCS-PL-V261B 2 P 2A Drain RCS-PL-V261C 2 P 2B Drain RCS-PL-V261D 2 S Pressure Boundary Valve Thermal Relief Isolation RNS-PL-V004A 2 S Pressure Boundary Valve Thermal Relief Isolation RNS-PL-V004B 2 S Pump A Suction Isolation RNS-PL-V005A 2 S Pump B Suction Isolation RNS-PL-V005B 2 S HX A Outlet Flow Control RNS-PL-V006A 2 S HX B Outlet Flow Control RNS-PL-V006B 2 S Pump A Discharge Isolation RNS-PL-V007A 2 S Pump B Discharge Isolation RNS-PL-V007B 2 S HX A Bypass Flow Control RNS-PL-V008A 2 S HX B Bypass Flow Control RNS-PL-V008B 2 S Discharge Containment Isolation Valve Test RNS-PL-V010 2 S Discharge Containment Isolation Valve Test Connection RNS-PL-V014 2 S Discharge Containment Penetration Isolation Valves Test RNS-PL-V016 2 S Discharge to IRWST Isolation RNS-PL-V024 2 S Discharge to CVS RNS-PL-V029 2 S Train A Discharge Flow Instrument Isolation RNS-PL-V031A 2 S Train B Discharge Flow Instrument Isolation RNS-PL-V031B 2 3I-61 Revision 1
AP1000 Description Tag Number Comment S Train A Discharge Flow Instrument Isolation RNS-PL-V032A 2 S Train B Discharge Flow Instrument Isolation RNS-PL-V032B 2 S Pump A Suction Pressure Instrument Isolation RNS-PL-V033A 2 S Pump B Suction Pressure Instrument Isolation RNS-PL-V033B 2 S Pump A Discharge Pressure Instrument Isolation RNS-PL-V034A 2 S Pump B Discharge Pressure Instrument Isolation RNS-PL-V034B 2 S Pump A Suction Piping Drain Isolation RNS-PL-V036A 2 S Pump B Suction Piping Drain Isolation RNS-PL-V036B 2 S HX A Channel Head Drain Isolation RNS-PL-V046A 2 S HX B Channel Head Drain Isolation RNS-PL-V046B 2 S Pump A Casing Drain Isolation RNS-PL-V050 2 S Pump B Casing Drain Isolation RNS-PL-V051 2 S Suction from SFP Isolation RNS-PL-V052 2 S Discharge to SFP Isolation RNS-PL-V053 2 S Suction from Cask Loading Pit Isolation Valve RNS-PL-V055 2 S Pump Suction to Cask Loading Pit Isolation RNS-PL-V056 2 S Train A Miniflow Isolation Valve RNS-PL-V057A 2 S Train B Miniflow Isolation Valve RNS-PL-V057B 2 S Pump Suction Containment Isolation Test Connection RNS-PL-V059 2 S Discharge to DVI Line A Drain RNS-PL-V066A 2 S Discharge to DVI Line B Drain RNS-PL-V066B 2 S Discharge to DVI Line A Drain RNS-PL-V067A 2 S Discharge to DVI Line B Drain RNS-PL-V067B 2 S Discharge to IRWST Drain RNS-PL-V068 2 19A Root Isolation Valve SFS-PL-V024A 2 19B Root Isolation Valve SFS-PL-V024B 2 3I-62 Revision 1
AP1000 Description Tag Number Comment 19C Root Isolation Valve SFS-PL-V024C 2 20 Root Isolation Valve SFS-PL-V028 2 Refueling Cavity Drain To SGS Compartment Isolation SFS-PL-V031 2 Refueling Cavity Suction Isolation SFS-PL-V032 2 Refueling Cavity Drain to Containment Sump Isolation SFS-PL-V033 2 Suction Line from IRWST Isolation SFS-PL-V039 2 Fuel Transfer Canal Suction Isolation SFS-PL-V040 2 Cask Loading Pit Suction Isolation SFS-PL-V041 2 CVS Makeup Reverse Flow Prevention SFS-PL-V043 2 Containment Penetration Test Connection SFS-PL-V048 2 Containment Penetration Test Connection Isolation SFS-PL-V056 2 Containment Isolation Valve V034 Test SFS-PL-V058 2 Containment Floodup Isolation Valve SFS-PL-V075 2 01 Root Isolation Valve SGS-PL-V001A 2 05 Root Isolation Valve SGS-PL-V001B 2 01 Root Isolation Valve SGS-PL-V002A 2 05 Root Isolation Valve SGS-PL-V002B 2 02 Root Isolation Valve SGS-PL-V003A 2 06 Root Isolation Valve SGS-PL-V003B 2 02 Root Isolation Valve SGS-PL-V004A 2 06 Root Isolation Valve SGS-PL-V004B 2 03 Root Isolation Valve SGS-PL-V005A 2 07 Root Isolation Valve SGS-PL-V005B 2 03 Root Isolation Valve SGS-PL-V006A 2 07 Root Isolation Valve SGS-PL-V006B 2 04 Root Isolation Valve SGS-PL-V007A 2 3I-63 Revision 1
AP1000 Description Tag Number Comment 08 Root Isolation Valve SGS-PL-V007B 2 04 Root Isolation Valve SGS-PL-V008A 2 08 Root Isolation Valve SGS-PL-V008B 2 11 Root Isolation Valve SGS-PL-V010A 2 13 Root Isolation Valve SGS-PL-V010B 2 11 Root Isolation Valve SGS-PL-V011A 2 13 Root Isolation Valve SGS-PL-V011B 2 12 Root Isolation Valve SGS-PL-V012A 2 14 Root Isolation Valve SGS-PL-V012B 2 12 Root Isolation Valve SGS-PL-V013A 2 14 Root Isolation Valve SGS-PL-V013B 2 21 Root Isolation Valve SGS-PL-V015A 2 23 Root Isolation Valve SGS-PL-V015B 2 20 Root Isolation Valve SGS-PL-V016A 2 22 Root Isolation Valve SGS-PL-V016B 2 21 Root Isolation Valve SGS-PL-V017A 2 23 Root Isolation Valve SGS-PL-V017B 2 20 Root Isolation Valve SGS-PL-V018A 2 22 Root Isolation Valve SGS-PL-V018B 2 n Steam Line Vent Isolation SGS-PL-V019A 2 n Steam Line Vent Isolation SGS-PL-V019B 2 30 Root Isolation Valve SGS-PL-V022A 2 34 Root Isolation Valve SGS-PL-V022B 2 31 Root Isolation Valve SGS-PL-V023A 2 35 Root Isolation Valve SGS-PL-V023B 2 32 Root Isolation Valve SGS-PL-V024A 2 3I-64 Revision 1
AP1000 Description Tag Number Comment 36 Root Isolation Valve SGS-PL-V024B 2 33 Root Isolation Valve SGS-PL-V025A 2 37 Root Isolation Valve SGS-PL-V025B 2 am Line 1 Nitrogen Supply Isolation SGS-PL-V038A 2 am Line 2 Nitrogen Supply Isolation SGS-PL-V038B 2 V Bypass Control Isolation SGS-PL-V042A 2 V Bypass Control Isolation SGS-PL-V042B 2 V Bypass Control Isolation SGS-PL-V043A 2 V Bypass Control Isolation SGS-PL-V043B 2 1 Condensate Pipe Drain Valve SGS-PL-V045A 2 2 Condensate Pipe Drain Valve SGS-PL-V045B 2 15 Root Isolation Valve SGS-PL-V046A 2 17 Root Isolation Valve SGS-PL-V046B 2 15 Root Isolation Valve SGS-PL-V047A 2 17 Root Isolation Valve SGS-PL-V047B 2 16 Root Isolation Valve SGS-PL-V048A 2 18 Root Isolation Valve SGS-PL-V048B 2 16 Root Isolation Valve SGS-PL-V049A 2 18 Root Isolation Valve SGS-PL-V049B 2 44 Root Isolation Valve SGS-PL-V050A 2 46 Root Isolation Valve SGS-PL-V050B 2 44 Root Isolation Valve SGS-PL-V051A 2 46 Root Isolation Valve SGS-PL-V051B 2 45 Root Isolation Valve SGS-PL-V052A 2 47 Root Isolation Valve SGS-PL-V052B 2 45 Root Isolation Valve SGS-PL-V053A 2 3I-65 Revision 1
AP1000 Description Tag Number Comment 47 Root Isolation Valve SGS-PL-V053B 2 62 Root Isolation Valve SGS-PL-V056A 2 63 Root Isolation Valve SGS-PL-V056B 2 n Feedwater Check SGS-PL-V058A 2 n Feedwater Check SGS-PL-V058B 2 55A Root Isolation Valve SGS-PL-V062A 2 56A Root Isolation Valve SGS-PL-V062B 2 55A Root Isolation Valve SGS-PL-V063A 2 56A Root Isolation Valve SGS-PL-V063B 2 55A Root Isolation Valve SGS-PL-V064A 2 56A Root Isolation Valve SGS-PL-V064B 2 55A Root Isolation Valve SGS-PL-V065A 2 56A Root Isolation Valve SGS-PL-V065B 2 1 Nitrogen Sparging Isolation SGS-PL-V084A 2 2 Nitrogen Sparging Isolation SGS-PL-V084B 2 ice Isolation Valve SGS-PL-V093A 2 ice Isolation Valve SGS-PL-V093B 2 ice Cleanout Line Isolation Valve SGS-PL-V094A 2 ice Cleanout Line Isolation Valve SGS-PL-V094B 2 ice Isolation Valve SGS-PL-V095A 2 ice Isolation Valve SGS-PL-V095B 2 am Line Condensate Drain Level Isolation Valve SGS-PL-V096A 2 am Line Condensate Drain Level Isolation Valve SGS-PL-V096B 2 am Line Condensate Drain Level Isolation Valve SGS-PL-V097A 2 am Line Condensate Drain Level Isolation Valve SGS-PL-V097B 2 tup Feedwater Check Valve SGS-PL-V256A 2 3I-66 Revision 1
AP1000 Description Tag Number Comment tup Feedwater Check Valve SGS-PL-V256B 2 Delivery Line Pressure Instrument Isolation Valve A VES-PL-V006A 2 Delivery Line Pressure Instrument Isolation Valve B VES-PL-V006B 2 porary Instrument Isolation Valve A VES-PL-V016 2 eted eted porary Instrument Isolation Valve B VES-PL-V020 2 Tank Isolation Valve A VES-PL-V024A 2 Tank Isolation Valve B VES-PL-V024B 2 Tank Isolation Valve A VES-PL-V025A 2 Tank Isolation Valve B VES-PL-V025B 2 ll Line Isolation Valve VES-PL-V038 2 Instrument Line Isolation Valve A VES-PL-V043A 2 Instrument Line Isolation Valve B VES-PL-V043B 2 tainment Isolation Test Connection VFS-PL-V008 2 tainment Isolation Test Connection VFS-PL-V012 2 tainment Isolation Test Connection VFS-PL-V015 2 n Equipment Hatch Test Connection VUS-PL-V015 2 ntenance Equipment Hatch Test Connection VUS-PL-V016 2 sonnel Hatch Test Connection VUS-PL-V017 2 sonnel Hatch Test Connection VUS-PL-V018 2 sonnel Hatch Test Connection VUS-PL-V019 2 sonnel Hatch Test Connection VUS-PL-V020 2 sonnel Hatch Test Connection VUS-PL-V021 2 sonnel Hatch Test Connection VUS-PL-V022 2 l Transfer Tube Test Connection VUS-PL-V023 2 3I-67 Revision 1
AP1000 Description Tag Number Comment ctrical Penetration Test Isolation Valve VUS-PL-V101 2 ctrical Penetration Test Isolation Valve VUS-PL-V102 2 ctrical Penetration Test Isolation Valve VUS-PL-V103 2 ctrical Penetration Test Isolation Valve VUS-PL-V104 2 ctrical Penetration Test Isolation Valve VUS-PL-V105 2 ctrical Penetration Test Isolation Valve VUS-PL-V106 2 ctrical Penetration Test Isolation Valve VUS-PL-V107 2 ctrical Penetration Test Isolation Valve VUS-PL-V108 2 ctrical Penetration Test Isolation Valve VUS-PL-V109 2 ctrical Penetration Test Isolation Valve VUS-PL-V110 2 ctrical Penetration Test Isolation Valve VUS-PL-V111 2 ctrical Penetration Test Isolation Valve VUS-PL-V112 2 ctrical Penetration Test Isolation Valve VUS-PL-V113 2 ctrical Penetration Test Isolation Valve VUS-PL-V114 2 ctrical Penetration Test Isolation Valve VUS-PL-V115 2 ctrical Penetration Test Isolation Valve VUS-PL-V116 2 ctrical Penetration Test Isolation Valve VUS-PL-V117 2 ctrical Penetration Test Isolation Valve VUS-PL-V118 2 ctrical Penetration Test Isolation Valve VUS-PL-V119 2 ctrical Penetration Test Isolation Valve VUS-PL-V120 2 ctrical Penetration Test Isolation Valve VUS-PL-V121 2 ctrical Penetration Test Isolation Valve VUS-PL-V122 2 ctrical Penetration Test Isolation Valve VUS-PL-V123 2 ctrical Penetration Test Isolation Valve VUS-PL-V124 2 ctrical Penetration Test Isolation Valve VUS-PL-V125 2 3I-68 Revision 1
AP1000 Description Tag Number Comment re Penetration Test Connection VUS-PL-V140 2 re Penetration Test Connection VUS-PL-V141 2 re Penetration Test Connection VUS-PL-V142 2 S Supply Containment Penetration IRC Test Connection/Vent VWS-PL-V424 2 S Return Containment Penetration ORC Test Connection/Vent VWS-PL-V425 2 t Exchangers mal Residual Heat Removal Heat Exchanger A RNS-ME-01A 3 mal Residual Heat Removal Heat Exchanger B RNS-ME-01B 3 ks nt Fuel Pool FHS-MT-01 3 l Transfer Canal FHS-MT-02 3 nt Fuel Cask Loading Pit FHS-MT-05 3 sive Containment Cooling Water Storage Tank PCS-MT-01 3 er Distribution Bucket PCS-MT-03 3 er Collection Troughs PCS-MT-04 3 sive RHR Heat Exchanger PXS-ME-01 3 umulator Tank A PXS-MT-01A 3 umulator Tank B PXS-MT-01B 3 e Makeup Tank A PXS-MT-02A 3 e Makeup Tank B PXS-MT-02B 3 ontainment Refueling Water Storage Tank PXS-MT-03 3 ergency Air Storage Tank 01 VES-MT-01 3 ergency Air Storage Tank 02 VES-MT-02 3 ergency Air Storage Tank 03 VES-MT-03 3 ergency Air Storage Tank 04 VES-MT-04 3 3I-69 Revision 1
AP1000 Description Tag Number Comment ergency Air Storage Tank 05 VES-MT-05 3 ergency Air Storage Tank 06 VES-MT-06 3 ergency Air Storage Tank 07 VES-MT-07 3 ergency Air Storage Tank 08 VES-MT-08 3 ergency Air Storage Tank 09 VES-MT-09 3 ergency Air Storage Tank 10 VES-MT-10 3 ergency Air Storage Tank 11 VES-MT-11 3 ergency Air Storage Tank 12 VES-MT-12 3 ergency Air Storage Tank 13 VES-MT-13 3 ergency Air Storage Tank 14 VES-MT-14 3 ergency Air Storage Tank 15 VES-MT-15 3 ergency Air Storage Tank 16 VES-MT-16 3 ergency Air Storage Tank 17 VES-MT-17 3 ergency Air Storage Tank 18 VES-MT-18 3 ergency Air Storage Tank 19 VES-MT-19 3 ergency Air Storage Tank 20 VES-MT-20 3 ergency Air Storage Tank 21 VES-MT-21 3 ergency Air Storage Tank 22 VES-MT-22 3 ergency Air Storage Tank 23 VES-MT-23 3 ergency Air Storage Tank 24 VES-MT-24 3 ergency Air Storage Tank 25 VES-MT-25 3 ergency Air Storage Tank 26 VES-MT-26 3 ergency Air Storage Tank 27 VES-MT-27 3 ergency Air Storage Tank 28 VES-MT-28 3 ergency Air Storage Tank 29 VES-MT-29 3 ergency Air Storage Tank 30 VES-MT-30 3 3I-70 Revision 1
AP1000 Description Tag Number Comment ergency Air Storage Tank 31 VES-MT-31 3 ergency Air Storage Tank 32 VES-MT-32 3 n Feed Pump A Status ECS-ES-3-XXX 4 n Feed Pump B Status ECS-ES-4-XXX 4 n Feed Pump C Status ECS-ES-5-XXX 4 s:
Rugged AP1000 safety-related equipment with no moving parts required in demonstrating functional operability during a seismic event is considered to be not sensitive to HRHF seismic loadings. Seismic qualification is based on the seismic loads associated with the mounting location of the safety-related equipment as a minimum. AP1000 CSDRS seismic loads at the mounting location of the safety-related equipment produces comparable or higher equipment stresses and deflections than the HRHF seismic loadings based on the work reported in APP-GW-GLR-115, Effect of High Frequency Seismic Content on SSCs. For rugged safety-related line-mounted equipment being qualified by test, seismic testing will be performed in compliance with IEEE Standard 382-1996 with a required input motion (RIM) curve extended to 64 Hz typically to a peak acceleration of 6g.
AP1000 safety-related valves are seismic qualified in accordance with ASME code for structural integrity to a maximum acceleration of 6g in all three principal orthogonal axes. AP1000 CSDRS seismic loads at the mounting location of the safety-related equipment produce comparable or higher equipment stresses and deflections than the HRHF seismic loadings based on the work reported in APP-GW-GLR-115, Effect of High Frequency Seismic Content on SSCs. For rugged safety-related line-mounted equipment being qualified by test, seismic testing will be performed in compliance with IEEE Standard 382-1996 with a required input motion (RIM) curve extended to 64 Hz typically to a peak acceleration of 6g.
Seismic qualification is based on structural integrity alone to the seismic loadings associated with the mounting location of the safety-related equipment as a minimum. AP1000 CSDRS seismic loads at the mounting location of the safety-related equipment produce comparable or higher equipment stresses and deflections than the HRHF seismic loadings based on the work reported in APP-GW-GLR-115, Effect of High Frequency Seismic Content on SSCs.
Seismic qualification is not required.
3I-71 Revision 1
3I-72 Revision 1 3I-73 Revision 1 1.00 HRHF GMRS 0.90 AP1000 CSDRS 0.80 0.70 Acceleration (g) 0.60 0.50 0.40 0.30 0.20 0.10 0.00 0.1 1 10 100 Frequency (Hz)
Figure 3I.1-1 Comparison of Horizontal AP1000 CSDRS and HRHF Envelope Response Spectra 3I-74 Revision 1
1.00 0.90 HRHF GMRS AP1000 CSDRS 0.80 0.70 Acceleration (g) 0.60 0.50 0.40 0.30 0.20 0.10 0.00 0.1 1 10 100 Frequency (Hz)
Figure 3I.1-2 Comparison of Vertical AP1000 CSDRS and HRHF Envelope Response Spectra 3I-75 Revision 1
1 0.10 Acceleraon (g) 0.01 Explanation AP1000 CSDRS Horizontal AP1000 HRHF Horizontal NI FIRS Horizontal 0.001 0.1 1 10 100 Frequency (Hz)
Figure 3I.1-201 Design Ground Motion Response Spectra - NI FIRS Horizontal 3I-76 Revision 1
1 0.10 Acceleraon (g) 0.01 Explanation AP1000 CSDRS s AP1000 HRHF s E/&/Z^s 0.001 0.1 1 10 100 Frequency (Hz)
Figure 3I.1-202 Design Ground Motion Response Spectra - NI FIRS Vertical 3I-77 Revision 1