ML17086A600
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B/B-UFSAR 3.2-1 REVISION 12 - DECEMBER 2008 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 3.2.1 Safety Classification Structures, systems, and components are classified for design purposes as either Safety Category I or Safety Category II. Piping and instrumentation diagrams (P&IDs) provided in this UFSAR show boundaries of classification, i.e., classification changes, on applicable drawings.
3.2.1.1 Safety Category I Those structures, systems, and components important to safety that are designed to remain functional in the event of the safe shutdown earthquake (SSE) and other design-basis events (including tornado, probable maximum flood, operating basis earthquake - OBE, missile impact, or accident internal to the plant) are designated as Safety Category I. This category includes those structures, systems, and components whose safety function is to retain their own integrity and/or not constitute a hazard to the safety function of other Safety Category I structures, systems, and components.
Safety Category I structures, systems, and components are those necessary to assure: a. the integrity of the reactor coolant pressure boundary, b. the capability to shut down the reactor and maintain it in a safe shutdown condition, or c. the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100 for accidents analyzed using TID-14844 and 10 CFR 50.67 for accidents analyzed using Regulatory Guide 1.183 (AST).
Safety Category I systems and components are not located within Safety Category II structures, except as noted in Table 3.2-1. Table 3.2-1 lists all plant structures and major components which are designated as Safety Category I. Systems or portions of systems, including piping, which are designated as Safety Category I are identified on the system P&IDs associated with that system. The division between Safety Category I and II portions of systems is in accordance with the intent of the requirements for seismic design classification.
Safety Category I systems or portions of systems and components meet the requirements of Appendix B to 10 CFR 50.
B/B-UFSAR 3.2-2 REVISION 5 - DECEMBER 1994 3.2.1.2 Safety Category II Those structures, systems, and components which are not designated as Safety Category I are designated as Safety Category II. This category has no public health or safety implication. Safety Category II structures, systems, and components are not specifically designed to remain functional in the event of the safe shutdown earthquake (SSE) or other design-basis events (including tornado, probable maximum flood, operating basis earthquake, missile impact, or an accident internal to the plant). A reasonable margin of safety is, however, considered in the design as dictated by local requirements. Many Safety Category II items in Category I buildings are supported with seismically designed supports. These items and their supports are not Safety Category I or Seismic Category I as defined by Regulatory Guide 1.29. Structures and major components not listed in Table 3.2-1 as Safety Category I are Safety Category II. Safety Category II systems or portions of systems and components do not follow the requirements of Appendix B to 10 CFR 50. The quality assurance standards for these systems and components follow normal industrial standards and any other requirements deemed necessary by the Licensee. 3.2.2 Quality Group Classification The quality group classification is followed in which five quality groups (A, B, C, D, and G) are identified.
The following data indicates the overall correspondence between safety categories and quality groups and the general boundaries of systems to be considered part of each quality group. QUALITY GROUP SAFETY CATEGORY GENERAL SYSTEM DESCRIPTION A I Reactor coolant pressure boundary and extensions thereof. B I Emergency core cooling, post-LOCA heat removal and cleanup, safe reactor shutdown and heat removal, portions of main steam and feedwater associated with containment isolation. C I Cooling water and auxiliary feedwater systems or portions thereof that are designed for emergency core cooling; postaccident containment B/B-UFSAR 3.2-3 REVISION 8 - DECEMBER 2000 QUALITY GROUP SAFETY CATEGORY GENERAL SYSTEM DESCRIPTION cooling; postaccident containment atmosphere cleanup; and residual heat removal from the reactor and spent fuel storage (including primary and secondary cooling systems). C I Cooling water and seal water systems or portions of these systems that provide support for other systems and components important to safety. C I Systems or portions of systems that are capable of being isolated from the reactor coolant pressure boundary during all modes of normal reactor operation by two valves which are either normally closed or capable of automatic closure. C I Portions of the radioactive waste and other systems whose postulated failure would release radioactive isotopes that would result in a calculated offsite dose in excess of 0.5 rem to the whole body or its equivalent part (refer to Table 3.2-1). D II Systems designed to ANSI B31.1.0 code criteria. G I Safety Category I piping and/or components, non-ASME or under the jurisdiction of other codes. Diesel generator skid-mounted components, originally constructed to Safety Category I, Quality Group C requirements and reclassified in FSAR Amendment 49, are included in this classification (Byron only).
B/B-UFSAR 3.2-4 REVISION 6 - DECEMBER 1996 Table 3.2-1 lists all the Safety Category I components and the major Safety Category II components with their respective quality group classifications. Table 3.2-2 identifies the applicable codes and standards used for each quality group. Components which are not assigned specific quality group classifications are designed in accordance with normal industrial standards and good engineering practices. Table 3.2-3 cross-references the ANS safety classifications to the classification designations established for Byron/Braidwood.
B/B-UFSAR 3.6-46 TABLE 3.6-1 SYSTEMS IMPORTANT TO PLANT SAFETY Auxiliary Feedwater Residual Heat Removal Component Cooling Essential Service Water Essential Service Water Makeup (Byron only) Chemical and Volume Control
____________________
Note: For given postulated events, further additional systems may be required (e.g., safety injection required for a LOCA). See Subsection 3.6.1.3.
B/B-UFSAR 3.6-47 REVISION 9 - DECEMBER 2002 TABLE 3.6-2 SYSTEMS WHICH CONTAIN HIGH OR MODERATE ENERGY FLUID DURING NORMAL OPERATION HIGH ENERGY SYSTEM ACRONYM MODERATE ENERGY Auxiliary Steam AS Boric Acid and Boron Recycle Chemical and Volume Control CV Boron Thermal Regeneration Main Feedwater FW Chemical and Volume Control Main Steam MS Chemical Feed Radioactive Waste Processing WX Component Cooling Reactor Coolant RC Containment Air Monitoring/Sampling Steam Generator Blowdown SD Diesel Fuel Oil Essential Service Water Pressurizer RY Fire Protection Safety Injection Accumulators SI H2, N2, and CO2 Systems Instrument and Service Air Nonessential Service Water Radioactive Waste Processing Residual Heat Removal Spent Fuel Pit Cooling Station Heating Steam Generator Blowdown Waste Gas
____________________ Note: 1. Systems shown are either totally or partially high/moderate energy during normal operation. High/moderate energy lines can be identified through the engineering controlled equipment/component database(s). 2. The above listing reflects criteria with regard to systems operated infrequently, such as the Residual Heat Removal System. 3. The auxiliary feedwater system is not utilized for normal startup and shutdown of the unit. It is classified as a moderate-energy system.
B/B-UFSAR 3.6-48 REVISION 4 - DECEMBER 1992 TABLE 3.6-3 ESSENTIAL SYSTEMS VERSUS TYPE OF POSTULATED PIPING BREAK TYPE POSTULATED PIPING BREAK ESSENTIAL SYSTEMS I Large Reactor Coolant Break Safety Injection (Piping Larger than 4 inches) Residual Heat Removal Chemical and Volume Control Containment Spray Auxiliary Feedwater Component Cooling Essential Service Water Essential Service Water Makeup (Byron only) II Small Reactor Coolant Break Safety Injection (Piping Equal to or Smaller Residual Heat Removal than 4 inches) Chemical and Volume Control Containment Spray Auxiliary Feedwater Component Cooling Essential Service Water Essential Service Water Makeup (Byron only) III Critical Secondary Side Break Safety Injection (Main steam, feedwater, or Residual Heat Removal steam generator blowdown Chemical and Volume Control inside containment) Containment Spray Auxiliary Feedwater Component Cooling Essential Service Water Essential Service Water Makeup (Byron only)
B/B-UFSAR 3.6-49 REVISION 4 - DECEMBER 1992 TABLE 3.6-3 (Cont'd) TYPE POSTULATED PIPING BREAK ESSENTIAL SYSTEMS IV Noncritical Secondary Residual Heat Removal Side Break (Main steam, Chemical and Volume Control feedwater, or steam generator Auxiliary Feedwater blowdown outside isolation Component Cooling valve room) Essential Service Water Essential Service Water Makeup (Byron only) V Breaks other than Types I Residual Heat Removal through IV Chemical and Volume Control Auxiliary Feedwater Component Cooling Essential Service Water Essential Service Water Makeup (Byron only)
B/B-UFSAR 3.6-50 REVISION 4 - DECEMBER 1992 TABLE 3.6-4 BREAK PROPAGATION LIMITS VERSUS TYPE OF POSTULATED PIPING FAILURE (In addition to not propagating to essential systems listed in Table 3.6-3, postulated piping breaks are further limited as follows:) TYPE POSTULATED PIPING BREAK CANNOT PROPAGATE TO I Large Reactor Coolant Break a) Secondary Side (Piping larger than 4 inches) b) Unaffected loops c) Over 20% more than the original break area within the affected loop II Small Reactor Coolant Break a) Secondary Side (Piping equal to or smaller b) Unaffected loops than 4 inches) c) Unaffected legs within the affected loop d) Over 12.5 in2 in total area within the affected leg* *The only exception to this rule is that any high head injection line break cannot propagate to any other high head injection line, even in the same leg.
B/B-UFSAR 3.6-51 REVISION 4 - DECEMBER 1992 TABLE 3.6-4 (Cont'd) TYPE POSTULATED PIPING BREAK CANNOT PROPAGATE TO III Critical Secondary Side Break a) Reactor Coolant (Main Steam, feedwater, or steam b) Secondary side lines from generator blowdown inside the unaffected steam generators, containment) the unaffected steam generators themselves, and their drain piping IV Noncritical Secondary Side a) Reactor Coolant Break (main steam, feedwater, b) Any other piping in the or steam generator blowdown main steam tunnel** outside isolation valve room) V Breaks other than Types I a) Reactor Coolant through IV b) Secondary Side **This requirement is a limiting condition from the pressurization design of the main steam tunnel.
B/B-UFSAR 3.6-52 TABLE 3.6-5
Deleted B/B-UFSAR 3.6-53 through 3.6-56 REVISION 7 - DECEMBER 1998
Pages 3.6-53 through 3.6-56 have been deleted intentionally.
B/B-UFSAR 3.6-57 TABLE 3.6-8 ENERGY BALANCE VS FINITE DIFFERENCE ANALYSIS* (30-inch pipe) RESTRAINT DISPLACEMENT (in.)
L1 L2 (in.) (in.) ENERGY BALANCE FINITE DIFFERENCE 99.54 66.34 3.36 3.01 149.30 66.34 5.17 4.93 431.21 66.34 6.41 6.76 87.06 58.04 3.14 2.15 145.01 58.04 4.43 4.59 435.30 58.04 5.91 5.85 99.51 49.75 2.76 2.39 447.77 49.75 5.15 5.31 Pipe OD = 30 in. F = 607.2 kips Pipe wall = 1.1619 in. R = 1062.6 kips Pipe weight = 34.7 lb/in. GAP = 4 in. Plastic moment used in Energy Balance analysis = 50349 in kips. Material properties used in finite difference analysis.
Modulus of elasticity = 26100 ksi Power law constants K = 90.277 ksi n = 0.1648 in = K()n. *A106 Grade B carbon steel pipe at 550°F.
B/B-UFSAR 3.6-58 TABLE 3.6-9 ENERGY BALANCE VS FINITE DIFFERENCE ANALYSIS* (12-inch pipe) RESTRAINT DISPLACEMENT (in.)
L1 L2 (in.) (in.) ENERGY BALANCE FINITE DIFFERENCE 81.51 54.4 4.10 3.46 122.28 54.4 5.39 5.58 380.53 54.4 7.93 6.75 82.47 47.1 3.28 3.10 223.82 47.1 6.55 6.79 388.74 47.1 7.06 6.20 Pipe OD = 12.75 in. F = 95.3 kips Pipe wall = 0.95607 R = 166.78 kips Pipe weight = 14.08 lb/in. GAP = 4 in.
Plastic moment used in Energy Balance analysis = 6908 in. kips.
Material properties used in finite difference analysis.
Modulus of elasticity = 26100 ksi Power law constants: K = 90.277 ksi n = 0.1648 n = K()n.
- A106 Grade B carbon steel pipe at 550°F.
B/B-UFSAR 3.6-59 REVISION 1 - DECEMBER 1989
Table 3.6-10 has been deleted intentionally.
B/B-UFSAR 3.6-60 TABLE 3.6-11 CALCULATED STRESS AND CUMULATIVE USAGE FACTORS FOR POSTULATED BREAK POINTS (For ASME Sec. III Class 1 Piping Systems) PIPING SYSTEM CALCULATED STRESS NORMAL & UPSET BREAK PLANT CONDITIONS 2.4(Sm) CUMULATIVE LINE NUMBER(S) ID (psi) (psi) USAGE FACTOR RCS BYPASS Loop 1
See stress analysis report for Subsystem RC-01 RCS BYPASS Loop 2 See stress analysis report for Subsystem RC-02 RCS BYPASS Loop 3
See stress analysis report for Subsystem RC-03 RCS BYPASS Loop 4
See stress analysis report for Subsystem RC-04
B/B-UFSAR 3.6-61 REVISION 1 - DECEMBER 1989 TABLE 3.6-11 (Cont'd) PIPING SYSTEM CALCULATED STRESS NORMAL & UPSET BREAK PLANT CONDITIONS 2.4(Sm) CUMULATIVE LINE NUMBER(S) ID (psi) (psi) USAGE FACTOR RCS SURGE LINE
See stress analysis report for Subsystem RY-05 SAFETY INJECTION Loop 1
See stress analysis report for Subsystem RH-02 SAFETY INJECTION Loop 2 See stress analysis report for Subsystem SI-11 SAFETY INJECTION Loop 3
See stress analysis report for Subsystem RH-02 SAFETY INJECTION Loop 4 See stress analysis report for Subsystem SI-10 RESIDUAL HEAT REMOVAL See stress analysis report for Subsystem RH-02
B/B-UFSAR 3.6-62 TABLE 3.6-12 CALCULATED STRESSES FOR POSTULATED BREAK POINTS (For ASME Sec. III Class 2&3 and ANSI B31.1 Piping Systems) PIPING SYSTEM CALCULATED STRESS NORMAL & UPSET BREAK PLANT CONDITIONS 0.8(1.2Sb+SA) LINE NUMBER(S) ID (psi) (psi) FEEDWATER Loop 1 See stress analysis report for Subsystem FW-02 FEEDWATER Loop 2 See stress analysis report for Subsystem FW-02 FEEDWATER Loop 3 See stress analysis report for Subsystem FW-04 FEEDWATER Loop 4 See stress analysis report for Subsystem FW-05 B/B-UFSAR 3.6-63 TABLE 3.6-12 (Cont'd) PIPING SYSTEM CALCULATED STRESS NORMAL & UPSET BREAK PLANT CONDITIONS 0.8(1.2Sb+SA) LINE NUMBER(S) ID (psi) (psi) FEEDWATER IN M.S. TUNNEL See stress analysis report for Subsystem FW-01 MAIN STEAM Loop 1 See stress analysis report for Subsystem MS-05 MAIN STEAM Loop 2 See stress analysis report for Subsystem MS-06 MAIN STEAM Loop 3 See stress analysis report for Subsystem MS-07 MAIN STEAM Loop 4 See stress analysis report for Subsystem MS-08 B/B-UFSAR 3.6-64 TABLE 3.6-12 (Cont'd) PIPING SYSTEM CALCULATED STRESS NORMAL & UPSET BREAK PLANT CONDITIONS 0.8(1.2Sb+SA) LINE NUMBER(S) ID (psi) (psi) MAIN STEAM IN M.S. TUNNEL See stress analysis report for Subsystem MS-01 SAFETY INJECTION Loop 1 See stress analysis report for Subsystem RH-02 SAFETY INJECTION Loop 2 See stress analysis report for Subsystem SI-13 SAFETY INJECTION Loop 3 See stress analysis report for Subsystem RH-02 SAFETY INJECTION Loop 4 See stress analysis report for Subsystem SI-14 B/B-UFSAR 3.6-65 REVISION 4 - DECEMBER 1992 TABLE 3.6-13 RESULTS OF DYNAMIC ANALYSES FOR POSTULATED PIPE BREAKS - RESTRAINED PIPE - INSIDE CONTAINMENT PIPING SYSTEM RESTRAINT INFORMATION PEAK TIP DYNAMIC ALLOWABLE POSTULATED Fimp Timp FFINAL GAP DISPLACEMENT DEFLECTION LOAD STRAIN STRAIN BREAK ID RESTRAINT ID (kips) (sec.) (kips) (inches) (inches) (inches) (kips) (in/in) (in/in) FEEDWATER* C55 FWR-1 176 .001 75 1.463 6.33 .179 631 .026 .5 C50 FWR-9 191 .008 142 2.642 5.85 1.314 390 .07 .08
C61 FWR-10 176 .001 75 1.526 4.88 .1964 513 .078 .5
C56 FWR-19 191 .008 142 2.00 3.84 1.03 414 .07 .08
- See Figures 3.6-47 and 3.6-48.
B/B-UFSAR 3.6-66 REVISION 7 - DECEMBER 1998 TABLE 3.6-13 (Cont'd) PIPING SYSTEM RESTRAINT INFORMATION PEAK TIP DYNAMIC ALLOWABLE POSTULATED Fimp Timp FFINAL GAP DISPLACEMENT DEFLECTION LOAD STRAIN STRAIN BREAK ID RESTRAINT ID (kips) (sec.) (kips) (inches) (inches) (inches) (kips) (in/in) (in/in) MAIN STEAM** C1 MSP-1 660 .003 460 6.38 14.60 4.4 1121 .054 .08
C9 (Unit 1) MSP-8 660 .003 460 2.18 14.52 8.413 842 .062 .08
C9 (Unit 2) MSP-8 660 .003 460 6.573 14.99 4.486 1122 .035 .08
C17 MSP-16 775 .003 539 2.57 8.13 3.45 1076 .05 .08
C25 MSP-23 775 .003 539 3.95 10.92 4.34 1071 .075 .08 ** See Figure 3.6-56 through 3.6-59 B/B-UFSAR 3.6-67 REVISION 1- DECEMBER 1989 TABLE 3.6-13 (Cont'd) PIPING SYSTEM RESTRAINT INFORMATION PEAK TIP DYNAMIC ALLOWABLE POSTULATED Fimp Timp FFINAL GAP DISPLACEMENT DEFLECTION LOAD STRAIN STRAIN BREAK ID RESTRAINT ID (kips) (sec.) (kips) (inches) (inches) (inches) (kips) (in/in) (in/in) SAFETY INJECTION*** C500 SI1R-10B 100 .0004 35 1.578 2.86 0.896 98 .077 .08
C501 SI1R-30 102 .0013 7 3.76 9.15 .16 145 .0074 .08
C518 SI3R-640A 107 .0004 58 1.814 3.96 2.033 121 .467 .5
C519 SI3R-655 102 .0015 7 1.10 6.20 2.1 95 .24 .5
C507 SI4R-15B 108 .0010 8 0.688 2.81 1.86 123 .485 .5 C507 SI4R-35 102 .0013 7 3.812 11.62 0.294 124 .033 .08
C513 SI9R-475B 108 .0010 9 0.53 6.90 6.18 74 .049 .08
C513 SI9R-495 102 .0015 9 3.168 8.45 1.299 125 .004 .08
- See Figures 3.6-88 through 3.6-91.
B/B-UFSAR 3.6-68 REVISION 1- DECEMBER 1989 TABLE 3.6-13 (Cont'd) PIPING SYSTEM RESTRAINT INFORMATION PEAK TIP DYNAMIC ALLOWABLE POSTULATED Fimp Timp FFINAL GAP DISPLACEMENT DEFLECTION LOAD STRAIN STRAIN BREAK ID RESTRAINT ID (kips) (sec.) (kips) (inches) (inches) (inches) (kips) (in/in) (in/in) RCS SURGE LINE* C400/C403 RY - 350 1.53 11.60 3.70 469 0.370 0.50
C400/C403 RY - 350 1.44 5.73 0.36 527 0.049 0.50
C400/C403 RY - 350 1.25 6.00 0.46 570 0.058 0.50
C400/C403 RY - 350 2.25 6.00 0.21 399 0.026 0.50 C400/C403 RY - 350 1.13 8.05 0.90 368 0.276 0.50
C400/C403 RY - 350 1.06 5.62 0.15 458 0.039 0.50
C400/C403 RY - 350 1.25 12.95 2.51 527 0.228 0.50
- See Figure 3.6-77 B/B-UFSAR 3.6-69 REVISION 1- DECEMBER 1989 TABLE 3.6-13 (Cont'd) PIPING SYSTEM RESTRAINT INFORMATION PEAK TIP DYNAMIC ALLOWABLE POSTULATED Fimp Timp FFINAL GAP DISPLACEMENT DEFLECTION LOAD STRAIN STRAIN BREAK ID RESTRAINT ID (kips) (sec.) (kips) (inches) (inches) (inches) (kips) (in/in) (in/in) RCS BYPASS* C144 RC1-1 87 .002 42 0.5 1.33 0.227 140 .0016 .08
C149 RC1-4 73 .010 15 3.48 17.57 2.01 74 .045 .08
C150 RC2-1 87 .002 42 0.5 1.19 .08 180 .0021 .08
C155 RC2-4 73 .010 15 3.712 17.66 1.254 78 .03 .08 C156 RC3-1 87 .002 42 0.5 1.24 .081 176 .003 .08 C161 RC3-4 73 .010 15 3.86 18.50 2.91 78 .003 .08
C162 RC4-1 87 .002 42 0.5 1.17 .079 180 .002 .08
C167 RC4-4 73 .010 15 3.75 15.87 2.85 76 .004 .08
- See Figures 3.6-64 through 3.6-67.
B/B-UFSAR 3.6-70 REVISION 1- DECEMBER 1989 TABLE 3.6-13 (Cont'd) PIPING SYSTEM RESTRAINT INFORMATION PEAK TIP DYNAMIC ALLOWABLE POSTULATED Fimp Timp FFINAL GAP DISPLACEMENT DEFLECTION LOAD STRAIN STRAIN BREAK ID RESTRAINT ID (kips) (sec.) (kips) (inches) (inches) (inches) (kips) (in/in) (in/in) RESIDUAL HEAT REMOVAL*
No Longer Applicable
NOTES:
(1) Restraints may be loaded by more than one break. Only the break that results in the most severe restraint loading is shown. (2) When FFINAL Fimp, FFINAL is used for the entire loading duration. (3) For tension member where A193B7 bar is used, the allowable strain = 0.08 in/in (see Subsection 3.6.2.3.2).
(4) For compression members where a honeycomb material is used, the allowable strain = 0.5 in/in (see Subsection 3.6.2.3.2).
B/B-UFSAR 3.6-71 REVISION 4 - DECEMBER 1992 TABLE 3.6-14 RESULTS OF DYNAMIC ANALYSES FOR POSTULATED PIPE BREAKS - RESTRAINED PIPE - OUTSIDE CONTAINMENT PIPING SYSTEM RESTRAINT INFORMATION PEAK TIP DYNAMIC ALLOWABLE POSTULATED Fimp Timp FFINAL GAP DISPLACEMENT DEFLECTION LOAD STRAIN STRAIN BREAK ID RESTRAINT ID (kips) (sec.) (kips) (inches) (inches) (inches) (kips) (in/in) (in/in) MAIN STEAM IN TUNNEL (1MSOCC-32-3/4) The results of the HELB analysis of the as-built condition of piping outside containment have indicated that pipe whip restraints outside containment are not required at Byron/Braidwood.
B/B-UFSAR
ATTACHMENT A3.6 SUBCOMPARTMENT PRESSURIZATION STUDIES OUTSIDE THE CONTAINMENT B/B-UFSAR A3.6-i TABLE OF CONTENTS PAGE A3.6.1 Introduction A3.6-1 A3.6.2 Basic Assumptions A3.6-3 A3.6.3 Analytical Model A3.6-4 A3.6.4 Summary of Results A3.6-4 A3.6.5 Conclusions A3.6-5
B/B-UFSAR A3.6-1 REVISION 10 - DECEMBER 2004 ATTACHMENT A3.6 - SUBCOMPARTMENT PRESSURIZATION STUDIES OUTSIDE THE CONTAINMENT A3.6.1 Introduction In accordance with the Nuclear Regulatory Commissions' Standard Review Plan, Subsection 6.2.1.2 (Subcompartment Analysis), a transient differential pressure response analysis is completed for all subcompartments containing high energy fluid lines. This attachment discusses the results of such analyses for the auxiliary building subcompartments.
If a high-energy fluid line is postulated to break within a subcompartment, the sudden discharge of this fluid will cause transient differential pressure across the walls of the subcompartment. Therefore, for a particular structural design, the results of a transient differential pressure and temperature response analysis should be an integral part of the structural design criteria.
Seven different subcompartments were analyzed (see Figures A3.6-1 through A3.6-25):
- a. recycle waste evaporator rooms elevation 346 feet 0 inch, auxiliary building basement; b. letdown reheat heat exchanger rooms and valve aisles at elevation 346 feet 0 inch, auxiliary building basement; c. blowdown condenser room at elevation 364 feet 0 inch, auxiliary building upper basement; d. letdown heat exchanger rooms and valve aisles at elevation 383 feet 0 inch, auxiliary building upper basement; e. auxiliary steamline piping tunnel at elevation 394 feet 0 inch, auxiliary building upper basement (Braidwood only);
- f. surface condenser room at elevation 401 feet 0 inch, auxiliary building ground floor (Braidwood only); and
- g. radwaste evaporator room at elevation 414 feet 0 inch, auxiliary building ground floor (Braidwood only). Pertinent high energy fluid lines, in which circumferential breaks were postulated, are listed in Figure A3.6-1 along with the initial operating conditions and computed choked or limited break mass discharge rates. At Byron, the steam supply to the auxiliary steamline piping tunnel, surface condenser rooms, and radwaste evaporator rooms has been permanently isolated in the turbine building by the installation of a blank plate.
B/B-UFSAR A3.6-1a REVISION 12 - DECEMBER 2008 At Braidwood, blank-off plates have been installed in the auxiliary steam piping in the Turbine Building isolating the steam from the subcompartments described in a, e, f and g above. Since the possibility exists that the blank-off plates could be removed in the future, the analysis for high energy line breaks in these subcompartments remains current and in place.
B/B-UFSAR A3.6-2 REVISION 10 - DECEMBER 2004 A3.6.1.1 Subcompartment Pressurization For the purpose of protecting subcompartments from overpressurization, the CV, AS, SD, WX, MS, and FW systems were traced through the auxiliary building and all subcompartments containing high energy line were identified. The most severe break in the subcompartment was analyzed. The main steam (MS) and feedwater (FW) systems are routed entirely in an enclosed tunnel in the auxiliary building. The limiting break in this tunnel is a main steamline break. Section C3.6 fully describes the subcompartment pressurization analysis of a break in this tunnel.
The remainder of the auxiliary building was surveyed level by level to identify all subcompartments which could be pressurized by high energy line breaks. Figures 3.6-100 through 3.6-104 identify all areas containing high energy lines. The identification of the limiting line in each zone is also included. The zone numbers do not correspond to environmental qualification zones (Section 3.11).
Figure 3.6-100 represents elevation 346 feet 0 inch. Zone 1, the recycle waste evaporator room, has been analyzed and the results are reported in this section. Zones 2 and 3, letdown reheat heat exchanger rooms and valve areas, have been analyzed and the results are reported in this section. The assessment in this section addressed Zone 3, the more limiting zone. Figure 3.6-101 represents elevation 364 feet 0 inch. Zones 5A, 5B, 6A, 6B, 7A, and 7B, the positive displacement and centrifugal charging pump rooms, contain high pressure, low temperature lines. Failure of these lines (normal temperature of 115°F) will not cause pressurization or increase temperatures. Pipe whip and impingement are considered. Zones 9A and 9B contain portions of the steam generator blowdown system. Control valves upstream of these lines limit the blowdown flow and prevent the postulated breaks from impacting plant design. Zones 8A and 8B, blowdown condenser rooms, have been analyzed and the results are included in Subsection A3.6.4. Zones 11 and 12, blowdown condenser rooms, have been analyzed and the results are reported in Subsection A3.6.4.
Figure 3.6-102 represents elevation 383 feet 0 inch. Zones 11A, 11B, 11C, and 11D, letdown heat exchanger rooms, have been analyzed and the results are reported in Subsection A3.6.4. Zone 13, the auxiliary steamline piping tunnel (Braidwood only), has been analyzed and the results are reported in Subsection A3.6.4. At Byron, a blank plate has been installed in the steam supply to the auxiliary steamline piping tunnel (Zone 13). Zones 12A and 12B are very similar to Zones 11A through 11D in break size and subcompartment size and, therefore, the existing B/B-UFSAR A3.6-3 REVISION 10 - DECEMBER 2004 results are adequate. Zones 10A and 10B are large areas with only small high energy lines. The impact of a break in these areas is discussed in Subsection 3.6.1.3.1. Figure 3.6-103 represents elevation 401 feet 0 inch. Zones 16A, 16B, and 16C, the surface condenser rooms (Braidwood only), have been analyzed and the results have been reported in Section A3.6. At Byron, a blank plate has been installed in the steam supply to the surface condenser rooms (Zones 16A, 16B, and 16C). Zone 14 is a large open area. The limiting high energy line break is a 2-inch auxiliary steamline. This event is discussed in Subsection 3.6.1.3.1.
Figure 3.6-104 represents elevation 426 feet 0 inch. Zones 18A, 18B, and 18C, radwaste evaporator rooms (Braidwood only), have been analyzed and the results are reported in Subsection A3.6.4. At Byron, a blank plate has been installed in the steam supply to the radwaste evaporator rooms (Zones 18A, 18B, and 18C). A3.6.2 Basic Assumptions Several assumptions were common to all analyses. They are the following: a. The free volume of each subcompartment was used as the effective control volume. b. Only one break of the specific high energy line is assumed for each subcompartment. c. With regards to Standard Review Plan (SRP), Section 6.2.1.2 (II.5a, III), the doors to the compartments were assumed to begin to open linearly at 0.5 psi and fully opened at 1.0 psi differential pressure. It was further assumed that the doors were of the destructive blow-off type so that they remain open once they become partially or completely open. d. Once the pipe break was initiated, the atmosphere in the subcompartment was homogeneously mixed by the WARLOC and COMPARE/MODl codes, keeping water as a fine mist with 100% liquid carryover. e. A multiplier of 0.6 was applied to the vent critical flow path calculation between nodes in accordance with the acceptance criteria of SRP Section 6.2.1.2 (II.6). f. The calculated transient differential pressure was multiplied by a 1.4 factor in accordance with the acceptance criteria of SRP Section 6.2.1.2 (II.7). (The multiplier of 1.4 was not applied for Cases c and f. since this calculation was performed after the completion of construction.) g. Moody's critical flow method with a multiplier of 1.0 as stated in the acceptance criteria of SRP Section B/B-UFSAR A3.6-4 REVISION 10 - DECEMBER 2004 6.2.1.2 (II.2) was used to compute the break mass discharge rate when a limited flow rate was not available. A circumferential break was postulated for each of the lines listed in Figure A3.6-1. h. There were no cutoff valves associated with the postulated broken high energy lines. i. In analyses 1, 2, 3, 4, and 6, the subcompartment door was the only vent flow path.
A3.6.3 Analytical Model The control volume (node) and vent flow path simulation mode are depicted in Figure A3.6-2. This model was utilized for all analyses except those associated with the postulated 16-inch auxiliary steamline break at elevation 394 feet 0 inch, elevation 401 feet 0 inch, and elevation 414 feet 0 inch. The transient differential pressure and temperature time histories in each subcompartment were determined with the Sargent & Lundy WARLOC computer program (09.8.038-4.0) and the NRC's COMPARE/MODl code. These codes are multicell thermal-hydraulic transient analysis codes capable of predicting the short-term and long-term containment and associated subcompartment pressure and temperature response to an accidental (or abnormal) transient such as a design-basis loss-of-coolant accident or high energy pipe line break.
Figures A3.6-3, A3.6-6, A3.6-9, A3.6-12, and A3.6-19 (Braidwood only) list the dimensions and initial conditions of the control volumes and flow paths used for each analysis as input to the WARLOC and COMPARE codes. A3.6.4 SUMMARY OF RESULTS Figures A3.6-2 through A3.6-14 and Figures A3.6-18 through A3.6-25 (Braidwood only) depict the initial input, dimensions, and computed pressure-temperature time histories for each analysis. Following are brief summaries: a. recycle waste evaporator room at elevation 346 feet 0 inch, peak transient differential pressure of 1.2 psid, maximum transient temperature of 295°F; b. letdown reheat heat exchanger room at elevation 346 feet 0 inch, peak transient differential pressure of 0.7 psid, maximum transient temperature of 210°F; c. blowdown condenser room at elevation 346 feet 0 inch, peak transient differential pressure of 2.54 psid, maximum transient temperature of 212°F;
B/B-UFSAR A3.6-5 REVISION 10 - DECEMBER 2004 d. letdown heat exchanger room at elevation 383 feet 0 inch, peak transient differential pressure of 0.7 psid, maximum transient temperature of 175°F; e. auxiliary steamline piping tunnel at elevation 394 feet 0 inch, peak transient differential pressure of 1.85 psid, maximum transient temperature of 300°F (Braidwood only), at Byron, a blank plate has been installed in the steam supply to this zone such that the high temperature and pressure described is no longer possible; f. surface condenser room at elevation 401 feet 0 inch, peak transient differential pressure of 1.25 psid, maximum transient temperature of 295°F (Braidwood only), at Byron, a blank plate has been installed in the steam supply to this zone such that the high temperature and pressure described is no longer possible; and
- g. radwaste evaporator room at elevation 414 feet 0 inch, peak transient differential pressure of 1.8 psid, maximum transient temperature of 300°F (Braidwood only), at Byron, a blank plate has been installed in the steam supply to this zone such that the high temperature and pressure described is no longer possible. A3.6.5 Conclusions All of the results are conservative with respect to the assumption of not having cutoff valves in the postulated broken lines. However, the structural designs, in every case, could accommodate the postulated differential pressure transient or were modified accordingly.
The structural integrity of all mentioned subcompartments can be maintained in the very unlikely event of a circumferential break of any of the high energy lines in the auxiliary building.
B/B-UFSAR
ATTACHMENT B3.6 SELECTION OF PIPE MATERIAL PROPERTIES FOR USE IN PIPE WHIP ANALYSIS B/B-UFSAR B3.6-1 Revision 9 - December 2002 ATTACHMENT B3.6 SELECTION OF PIPE MATERIAL PROPERTIES FOR USE IN PIPE WHIP ANALYSIS A substantial amount of elevated temperature test data for A106 Grade B carbon steel is given in Reference 1. Material property values based on this data are used. Since little test data is available for TP304, TP304L, and TP316 stainless steels ASME Code specified values are used with the realization that they are very conservative.
The power law stress strain relationship is used for all steels. ()()nK= (1)
The effect of strain rate in carbon steels is accounted for (as suggested in Reference 2) by modifying Equation 1 as follows: ()nK/1D1,+= (2) Where D = 40.4 sec-1 = 5 This modification has been widely used - see for example References 2 and 3. For stainless steels the effect of strain rate is less pronounced (Reference 4) so that the use of a 10% increase in yield and ultimate strengths is used. A106 Grade B Material Properties at 600°F The results of tests on 71 specimens, in the temperature range of interest, are given in Reference 1. Twenty were tested at 600°F, 22 at room temperature, and the rest at temperatures between 200°F and 585°F. Yield stress (1) was shown to decrease - ultimate stress to increase - with increasing temperature. The minimum yield stress of any of the 71 specimens tested was 31.7 ksi, the average for the 20 tested at 600°F was 36.28 ksi with a sigma value of 3.67 ksi. The minimum ultimate stress value for all specimens was 64.4 ksi, the average for the 22 tested at room temperature 71.79 ksi with a sigma value of 4.72 ksi. The strength coefficient K and the hardening exponent n can be evaluated from the equations.
B/B-UFSAR B3.6-2 Revision 9 - December 2002 ()n002.Ky= nKnu= (3) Values of K and n obtained in this way are given in Table 1 below:
Yield Stress (Ksi) Ultimate Stress (Ksi) K (Ksi) n Minimum 31.7 64.4 86.445 0.16142 Mean-Sigma 32.61 67.08 90.204 0.16371 Mean 36.28 71.79 95.971 0.15653 Table 1 Material Properties A106 GrB at 600°F based on data from Reference 2. The mean-sigma values of K=90.204 Ksi and n = 0.16371, or more conservative values, are used for all temperatures 600°F and below.
___________________ (1) 0.2% offset values were measured
References:
- 1. R. J. Eiber, et al. Investigation of the Initiation and Extent of Ductile Pipe Rupture, Battelle Memorial Institute, Report BMI-1866, July 1969 2. S. R. Bodner and P. S. Symonds, "Experimental and Theoretical Investigation of the Plastic Deformation of Cantilever Beams Subjected to Impulsive Loading", JAM, December 1962 3. J. C. Anderson and A. K. Singh, "Inelastic Response of Nuclear Piping Subjected to Rupture Forces," ASME Paper No. 75-PVP-21 4. C. Albertini and M. Montagnani, "Wave Propagation Effects in Effects in Dynamic Loadings," Nuclear Engineering and Design, 37 115-124, 1976.
B/B-UFSAR
ATTACHMENT C3.6 MAIN STEAMLINE BREAK IN MAIN STEAM TUNNEL B/B-UFSAR C3.6-i REVISION 7 - DECEMBER 1998 TABLE OF CONTENTS PAGE I. Introduction C3.6-1 II. Analysis C3.6-2 A. Description of the Computer Code C3.6-2 B. Simulation of the System C3.6-2 1. Assumptions C3.6-2 2. Analytical Model C3.6-3 III. Results and Conclusions C3.6-3 A. Unit 1 Results C3.6-3 1. Pipe Break in the Main Steam Tunnel C3.6-3 2. Pipe Break in Valve Room C3.6-4 B. Unit 1 Conclusions C3.6-4 C. Unit 2 Results C3.6-4 1. Pipe Break in the Main Steam Tunnel C3.6-4 2. Pipe Break in Valve Room C3.6-4 D. Unit 2 Conclusions C3.6-5 IV. References C3.6-5
B/B-UFSAR C3.6-ii REVISION 7 - DECEMBER 1998 LIST OF TABLES NUMBER TITLE PAGE 1 Unit 1 Subcompartment Nodal Description C3.6-6 2 Unit 2 Subcompartment Nodal Description C3.6-9 3 Subcompartment Vent Flow Path Description C3.6-12 4 Unit 1 Blowdown Rates and Enthalpy for Main Steamline Break C3.6-17 5 Unit 2 Blowdown Rates and Enthalpy for Main Steamline Break C3.6-20 6 Summary of Unit 1 Peak Pressures Between Valve Rooms and Main Steam Tunnel C3.6-21
B/B-UFSAR C3.6-iii REVISION 7 - DECEMBER 1998 LIST OF FIGURES NUMBER TITLE C3.6-1 Main Steam Tunnel View Plan C3.6-2 Nodalization Schematic C3.6-3 Unit 1-Differential Pressure vs. Time, Main Steam Line Break in Lower Second Quadrant Valve Room (Node 5) (Volumes 2, 3, 4, and 5) C3.6-4 Unit 1-Differential Pressure vs. Time, Main Steam Line Break in Second Quadrant Tunnel (Node 6) and Main Steam Line Break in Second and First Quadrant Tunnel (Node 7) C3.6-5 Unit 1-Differential Pressure vs. Time, Main Steam Line Break in First Quadrant Tunnel (Node 8) and Main Steam Line Break in First Quadrant Tunnel (Node 13) C3.6-6 Unit 1-Differential Pressure vs. Time, Main Steam Line Break in First Quadrant Tunnel (Node 14) C3.6-7 Unit 2 Differential Pressure vs. Time for Nodes 2, 3, 4, and 5 (Break in Node 5) C3.6-8 Unit 2 Differential Pressure vs. Time for Nodes 6, 7, 8, 13, and 14 (Break in Node 5) C3.6-9 Unit 2 Differential Pressure vs. Time for Nodes 9, 10, 11, and 12 (Break in Node 5) C3.6-10 Unit 2 Differential Pressure vs. Time for Nodes 2, 3, 4, and 5 (Break in Node 6) C3.6-11 Unit 2 Differential Pressure vs. Time for Nodes 6, 7, 8, 13, and 14 (Break in Node 6) C3.6-12 Unit 2 Differential Pressure vs. Time for Nodes 9, 10, 11, and 12 (Break in Node 6) C3.6-13 Unit 2 Differential Pressure vs. Time for Nodes 2, 3, 4, and 5 (Break in Node 7) C3.6-14 Unit 2 Differential Pressure vs. Time for Nodes 6, 7, 8, 13, and 14 (Break in Node 7) C3.6-15 Unit 2 Differential Pressure vs. Time for Nodes 9, 10, 11, and 12 (Break in Node 7) C3.6-16 Unit 2 Differential Pressure vs. Time for Nodes 2, 3, 4, and 5 (Break in Node 8) C3.6-17 Unit 2 Differential Pressure vs. Time for Nodes 6, 7, 8, 13, and 14 (Break in Node 8) C3.6-18 Unit 2 Differential Pressure vs. Time for Nodes 9, 10, 11, and 12 (Break in Node 8)
B/B-UFSAR C3.6-1 REVISION 7 - DECEMBER 1998 ATTACHMENT C3.6 - MAIN STEAMLINE BREAK IN MAIN STEAM TUNNEL I. Introduction One of the design criteria for the main steam tunnel and valve room subcompartments is to retain functional integrity indefinitely, that is, to have the capability of withstanding peak transient differential pressures under a postulated accident mode.
It was the purpose of this study to determine the transient pressure response in the Unit 1 and Unit 2 main steam tunnels and the associated safety valve rooms in the first and second quadrants at the time of a sudden and complete circumferential failure of a main steamline. Six break locations for Unit 1 and four for Unit 2 were considered. The common locations are the lower valve room just downstream of the isolation valve; the main steam tunnel just outside the valve room in the first quadrant; the main steam tunnel between the first and second quadrants; and the main steam tunnel just outside the valve room in the second quadrant. Additionally, Unit 1 was evaluated for two breaks in the first quadrant between the valve room and the turbine building opening. Qualification tests have been conducted for the components in the safety valve house. The components include the main steam and main feedwater isolation valves, the main steam power-operated relief valve, and the main steam safety valves. These tests conservatively applied aging, radiation, seismic, and worst case environmental (temperature, pressure, and humidity) loading to the components, and showed that loss of function did not occur. The portion of the main steam and main feedwater pipe in the tunnel between the safety valve house and the turbine building meets the guidelines of Branch Technical Position APSCB3-1. A special pipe whip restraint is located around each pipe as it passes through the wall separating the isolation valve room from the main steam tunnel. This restraint limits the amount of strain that can be transmitted to the isolation valves from any pipe break in the tunnel to a level which will not interfere with the proper functioning of the isolation valves.
The safety valve house, the steam tunnel, and the compartment between the containment and the safety valve house all have the same basis for design. These compartments have been designed for pressurization, impingement, and temperature as specified in Table 3.8-10, load combinations 8, 13, and 14.
B/B-UFSAR C3.6-2 REVISION 7 - DECEMBER 1998 An assumed pipe crack or break in the tunnel, isolation valve room , or safety valve house cannot cause structural failure. The subcompartment pressurization analysis is included in this attachment. The methods used to calculate the pressure buildup in subcompartments outside the containment are the same as those used for subcompartments inside the containment. II. Analysis A. Description of the Computer Code The analysis was carried out by using the RELAP code, which is a multicell thermal-hydraulic transient analysis computer program. The basis for the computer code is a network of fluid control volumes (fluid nodes) and fluid paths (interconnecting control volumes) for which the conservation equations of mass, momentum, and energy are solved in space and time. Superimposed on the network are computer subroutines which permit physical modeling of the reactor system, the containment, plant subcompartments, safeguard fluid systems and the pipe break flow. B. Simulation of the System 1. Assumptions The following are the major assumptions used in this study: a. The initial conditions in the steam tunnel and the valve rooms are 14.7 psia of pressure at a temperature of 90°F and a relative humidity of 30%.
- b. Only one break occurred per analysis. c. The Moody choked-flow calculation was used with a multiplier of 0.6 as required by the NRC for choked-flow check between nodes.
- d. Homogeneous fine mist for the steam/water-air system in the control volumes with complete liquid carryover was used to produce a conservative solution. e. The length of a flow path connecting any two control volumes was taken as the distance between the centroids of these volumes.
- f. The area of a flow path is the effective area (i.e., the cross-sectional area of the path excluding areas occupied by grating, pipes, louvers, etc.).
B/B-UFSAR C3.6-3 REVISION 7 - DECEMBER 1998 g. Mass and energy release rate for a postulated main steamline break is included in Tables 4 and 5 for Unit 1 and Unit 2 respectively. h. The doors and HVAC louvers/panels in the upper chambers of the valve rooms are initially assumed closed or intact. A differential pressure equal to 1.5 psi will blow open the doors and panels to atmosphere.
- 2. Analytical Model To determine the transient pressures and temperatures in the main steam tunnel and the safety valve rooms after a sudden failure of a main steamline, the main steam tunnel was simulated by five control volumes connected by flow paths. The area of each flow path is equivalent to the net area of the steam tunnel. The subcompartments of the valve room in each quadrant were represented by four control volumes connected by flow paths. The area of each flow path was equivalent to the total vent areas between subcompartments.
Figures 1 and 2 depict a plan of the system and the flow diagram of the analytical model used in the study, respectively.
Tables 1, 2 and 3 give the dimensions of the control volumes and flow paths, while Tables 4 and 5 show the blowdown rates and properties versus time from a postulated main steamline break as provided by Framatome Technologies International for Unit 1 (Reference 2) and Westinghouse for Unit 2, respectively. III. Results and Conclusions A. Unit 1 Results The peak nodal differential pressures, which represent the difference between steam tunnel/valve room nodes and the surrounding areas, are presented in Table 1 and Figures C3.6-3 through C3.6-6.
The peak differential pressures across internal walls and floors are shown in Table 6 and Figures C3.6-7 and C3.6-8.
- 1. Pipe Break in the Main Steam Tunnel Five locations of a postulated main steamline break were considered in the main steam tunnel. The first and second locations were just outside the valve rooms in the first and second quadrants (Nodes 6 and 8), respectively. The third
B/B-UFSAR C3.6-4 REVISION 7 - DECEMBER 1998 location was in the steam tunnel between the first and second quadrants (Node 7). The last two break locations are in the tunnel leading to the entrance to the turbine building (Nodes 13 and 14).
Figures C3.6-4 through C3.6-6 show the differential pressures in the control volumes directly affected by the line breaks in the tunnel. 2. Pipe Break in Valve Room A pipe break in the lower chamber of the valve room in the second quadrant (Node 5) was evaluated to provide the most conservative differential pressure. Figure C3.6-3 shows the differential pressures in the control volumes directly affected by the line break in the valve room.
Tables 1 and 6 provide a summary of the peak pressures used in the qualification of the structure (References 3 and 4). B. Unit 1 Conclusions The integrity of the Unit 1 main steam tunnel and valve rooms in both the first and second quadrants can be maintained during a postulated main steamline break. These differential pressures, modified by dynamic load factors, were used to qualify the subject walls for the tunnels (Reference 3) and valve rooms (Reference 4).
C. Unit 2 Results
- 1. Pipe Break in the Main Steam Tunnel Three locations of a postulated main steamline break were considered in the main steam tunnel. The first and second locations were just outside the valve rooms in the first and second quadrants (Nodes 6 and 8), respectively. The third location was in the steam tunnel between the first and second quadrants (Node 7). Figures 10 through 18 show the differential pressures in the control volumes directly affected by the line break.
- 2. Pipe Break in Valve Room A pipe break in the lower chamber of the valve room in the second quadrant (Node 5) was considered to give the most conservative differential pressure.
Figures 7 through 9 show the differential pressures in the affected control volumes after a line break.
B/B-UFSAR C3.6-5 REVISION 7 - DECEMBER 1998 Table 2 gives a summary of the peak pressures to the valves used in the design of the structure. D. Unit 2 Conclusions The integrity of the main steam tunnel, the auxiliary feedwater tunnel, and the valve rooms in both the first and second quadrants can be maintained during a postulated main steamline break.
IV. References
- 1. Calculation 3C8-0282-001, Revision 3. 2. NDIT 960136, "Steam Generator Replacement Project: Transmittal of Steam Line Break Mass and Energy for Steam Tunnel Pressure Analysis," dated September 16, 1996. 3. Calculation 5.6.1-BYR96-233/5.6.1-BRW-96-604, Revision 0. 4. Calculation 5.6.3-BYR96-234/5.6.3-BRW-96-608, Revision 0.
B/B-UFSAR C3.6-6 REVISION 7 - DECEMBER 1998 TABLE 1 UNIT 1 SUBCOMPARTMENT NODAL DESCRIPTION MAIN STEAM LINE BREAK IN UNIT 1 MAIN STEAM TUNNEL OR VALVE ROOMS DBA BREAK CONDITIONS CALC. DESIGN CROSS- BREAK PEAK PEAK SECTIONAL INITIAL CONDITIONS LOC. BREAK PRESS PRESS DESIGN VOLUME HEIGHT AREA VOLUME TEMP. PRESS. HUMID.*VOL. BREAK AREA BREAK DIFF. DIFF. MARGIN NO. DESCRIPTION ft ft2 ft3 °F psia % NO. LINE ft2 TYPE psid psid % 1 Atmosphere 5x103 1x103 107 90 14.7 30 - - - - - - - 2 2nd Quad- 12.33 133.25 4895.790 14.7 30 5 Main 1.4 Double- 15.3 26.2 71 rant Upper Steam ended Valve Guillo-Chamber tine 3 2nd Quad- 12.33 183.7 4895.790 14.7 30 5 Main 1.4 Double- 15.3 26.2 71 rant Upper Steam ended Valve Guillo-Chamber tine 4 2nd Quad- 24.00 213.0 6007.090 14.7 30 5 Main 1.4 Double- 17.7 26.2 48 rant Lower Steam ended Valve Guillo-Chamber tine 5 2nd Quad- 24.00 213.0 6007.090 14.7 30 5 Main 1.4 Double- 17.7 28.6 62 rant Valve Steam ended Chamber Guillo-tine
- Relative humidity.
B/B-UFSAR C3.6-7 REVISION 7 - DECEMBER 1998 TABLE 1 (Cont'd) DBA BREAK CONDITIONS CALC. DESIGN CROSS- BREAK PEAK PEAK SECTIONAL INITIAL CONDITIONS LOC. BREAK PRESS PRESS DESIGN VOLUME HEIGHT AREA VOLUME TEMP. PRESS. HUMID.*VOL. BREAK AREA BREAK DIFF. DIFF. MARGIN NO. DESCRIPTION ft ft2 ft3 °F psia % NO. LINE ft2 TYPE psid psid % 6 2nd Quad- 19.00 317.0 13695.090 14.7 30 6 Main 1.4 Double- 16.0 26.5 66 rant Main Steam ended Steam Guillo-Tunnel tine 7 Main Steam 19.00 203.00 34865.090 14.7 30 7 Main 1.4 Double- 15.2 21.3 40 Tunnel Steam ended Guillo- tine 8 1st Quad- 20.00 432.0 35016.090 14.7 30 8 Main 1.4 Double- 8.3 11.2 35 drant Steam Steam ended Tunnel Guillo-tine 9 1st Quad- 12.33 133.25 4895.790 14.7 30 5** Main 1.4 Double- 15.3 26.2 71 rant Upper Steam ended Valve Guillo-Chamber tine 10 1st Quad- 12.33 183.7 4895.790 14.7 30 5** Main 1.4 Double- 15.3 26.2 71 rant Upper Steam ended Valve Guillo- Chamber tine
- Relative humidity.
- Differential pressures calculated for 1st quadrant valve chambers are also applicable to corresponding 2nd quadrant valve chambers.
B/B-UFSAR C3.6-8 REVISION 7 - DECEMBER 1998 TABLE 1 (Cont'd) DBA BREAK CONDITIONS CALC. DESIGN CROSS- BREAK PEAK PEAK SECTIONAL INITIAL CONDITIONS LOC. BREAK PRESS PRESS DESIGN VOLUME HEIGHT AREA VOLUME TEMP. PRESS. HUMID.*VOL. BREAK AREA BREAK DIFF. DIFF. MARGIN NO. DESCRIPTION ft ft2 ft3 °F psia % NO. LINE ft2 TYPE psid psid % 11 1st Quad- 24.00 213.0 6007.090 14.7 30 5** Main 1.4 Double- 17.7 28.6 62 rant Valve Steam ended Chamber Guillo-tine 12 1st Quadrant 24.00 213.0 6007.090 14.7 30 5** Main 1.4 Double- 17.7 28.6 62 Lower Valve Steam ended Chamber Guillo- tine 13 1st Quad- 29.00 432.0 17388.490 14.7 30 13 Main 1.4 Double- 10.7 13.2 23 rant Main Steam ended Steam Guillo-Tunnel tine 14 1st Quad- 19.00 280.0 13529.990 14.7 30 14 Main 1.4 Double- 11.3 *** *** rant Main Steam ended Steam Guillo-Tunnel tine
- Relative humidity.
- Differential pressures calculated for 1st quadrant valve chambers are also applicable to corresponding 2nd quadrant valve chambers.
- The calculated peak differential pressure in Node 14 has been evaluated to be within plant design basis code allowable stresses (Reference 3).
B/B-UFSAR C3.6-9 REVISION 7 - DECEMBER 1998 TABLE 2 UNIT 2 SUBCOMPARTMENT NODAL DESCRIPTION MAIN STEAM LINE BREAK IN UNIT 2 MAIN STEAM TUNNEL OR VALVE ROOMS DBA BREAK CONDITIONS CALC. DESIGN CROSS- BREAK PEAK PEAK SECTIONAL INITIAL CONDITIONS LOC. BREAK PRESS PRESS DESIGN VOLUME HEIGHT AREA VOLUME TEMP. PRESS. HUMID.*VOL. BREAK AREA BREAK DIFF. DIFF. MARGIN NO. DESCRIPTION ft ft2 ft3 °F psia % NO. LINE ft2 TYPE psid psid % 1 Atmosphere 5x103 1x103 107 90 14.7 30 - - - - - - - 2 2nd Quad- 12.33 133.25 4895.790 14.7 30 5 Main 1.4 Double- 17.4 26.2 51 rant Upper Steam ended Valve Guillo-Chamber tine 3 2nd Quad- 12.33 183.7 4895.790 14.7 30 5 Main 1.4 Double- 17.4 26.2 51 rant Upper Steam ended Valve Guillo-Chamber tine 4 2nd Quad- 24.00 213.0 6007.090 14.7 30 5 Main 1.4 Double- 17.4 26.2 51 rant Lower Steam ended Valve Guillo-Chamber tine 5 2nd Quad- 24.00 213.0 6007.090 14.7 30 5 Main 1.4 Double- 19.7 28.6 45 rant Valve Steam ended Chamber Guillo- tine
- Relative humidity.
B/B-UFSAR C3.6-10 REVISION 7 - DECEMBER 1998 TABLE 2 (Cont'd) DBA BREAK CONDITIONS CALC. DESIGN CROSS- BREAK PEAK PEAK SECTIONAL INITIAL CONDITIONS LOC. BREAK PRESS PRESS DESIGN VOLUME HEIGHT AREA VOLUME TEMP. PRESS. HUMID.*VOL. BREAK AREA BREAK DIFF. DIFF. MARGIN NO. DESCRIPTION ft ft2 ft3 °F psia % NO. LINE ft2 TYPE psid psid % 6 2nd Quad- 19.00 317.0 13695.090 14.7 30 6 Main 1.4 Double- 16.4 26.5 61 rant Main Steam ended Steam Guillo-Tunnel tine 7 Main Steam 19.00 203.00 34865.090 14.7 30 7 Main 1.4 Double- 15.5 21.3 38 Tunnel Steam ended Guillo- tine 8 1st Quad- 20.00 432.0 35016.090 14.7 30 8 Main 1.4 Double- 8.8 11.2 28 drant Steam Steam ended Tunnel Guillo-tine 9 1st Quad- 12.33 133.25 4895.790 14.7 30 5** Main 1.4 Double- 17.4 26.2 51 rant Upper Steam ended Valve Guillo-Chamber tine 10 1st Quad- 12.33 183.7 4895.790 14.7 30 5** Main 1.4 Double- 17.4 26.2 51 rant Upper Steam ended Valve Guillo- Chamber tine
- Relative humidity. ** Differential pressures calculated for 1st quadrant valve chambers are also applicable to corresponding 2nd quadrant valve chambers.
B/B-UFSAR C3.6-11 REVISION 7 - DECEMBER 1998 TABLE 2 (Cont'd) DBA BREAK CONDITIONS CALC. DESIGN CROSS- BREAK PEAK PEAK SECTIONAL INITIAL CONDITIONS LOC. BREAK PRESS PRESS DESIGN VOLUME HEIGHT AREA VOLUME TEMP. PRESS. HUMID.*VOL. BREAK AREA BREAK DIFF. DIFF. MARGIN NO. DESCRIPTION ft ft2 ft3 °F psia % NO. LINE ft2 TYPE psid psid % 11 1st Quad- 24.00 213.0 6007.090 14.7 30 5** Main 1.4 Double- 19.7 28.6 45 rant Valve Steam ended Chamber Guillo-tine
12 1st Quadrant 24.00 213.0 6007.090 14.7 30 5** Main 1.4 Double- 19.7 28.6 45 Lower Valve Steam ended Chamber Guillo- tine
13 1st Quad- 29.00 432.0 17388.490 14.7 30 8 Main 1.4 Double- 8.9 13.2 48 rant Main Steam ended Steam Guillo-Tunnel tine
14 1st Quad- 19.00 280.0 13529.990 14.7 30 8 Main 1.4 Double- 5.9 10.3 75 rant Main Steam ended Steam Guillo-Tunnel tine
- Relative humidity. ** Differential pressures calculated for 1st quadrant valve chambers are also applicable to corresponding 2nd quadrant valve chambers.
B/B-UFSAR C3.6-12 REVISION 7 - DECEMBER 1998 TABLE 3 SUBCOMPARTMENT VENT PATH DESCRIPTION MAIN STEAM LINE BREAK IN MAIN STEAM TUNNEL OR VALVE ROOM FROM TO DESCRIPTION VENT VOL. VOL. OF HYDRAULIC HEAD LOSS, K PATH NODE NODE VENT PATH FLOW* AREAf LENGTHf DIAMETER FRICTIONTURNING EXPAN- CONTRAC- NO. NO. NO. CHOKED UNCHOKED ft2 ft ft K, ft/d LOSS, K SION, K TION, K TOTAL 1 14 1 Main Steam Tunnel to Turbine 270.8 28.0 13.9 - - - - 1.0656 Building Unchoked 2 2 5 2nd Quadrant Upper Valve Chamber 121.0 16.4 5.7 - - - - 1.5207 to Lower Valve Chamber Unchoked 3 3 4 2nd Quadrant Upper Valve Chamber 121.0 16.4 5.7 - - - - 1.5207 to Lower Valve Chamber Unchoked 4 6 4 2nd Quadrant Lower Valve Chamber 73.0 17.2 4.5 - - - - 1.5685 to Main Steam Tunnel Unchoked 5 6 5 2nd Quadrant Lower Valve Chamber 73.0 17.2 4.5 - - - - 1.5685 to Main Steam Tunnel Choked (5)
- See Figures 1 and 2. f Length/area is the inertial term input directly into RELAP4/MOD5.
Number in parentheses indicates the volume number of the break location which caused choke flow in the vent.
For break locations not indicated, unchoked flow had occurred for the vent.
B/B-UFSAR C3.6-13 REVISION 7 - DECEMBER 1998 TABLE 3 (Cont'd) FROM TO DESCRIPTION VENT VOL. VOL. OF HYDRAULIC HEAD LOSS, K PATH NODE NODE VENT PATH FLOW* AREAf LENGTHf DIAMETER FRICTIONTURNING EXPAN- CONTRAC- NO. NO. NO. CHOKED UNCHOKED ft2 ft ft K, ft/d LOSS, K SION, K TION, K TOTAL 6 5 4 Openings Between 2nd Quadrant 100.0 16.1 7.1 - - - - 1.5071 Lower Valve Chambers Unchoked 7 6 7 2nd Quadrant Main Steam Tunnel 199.8 102.0 13.3 - - - - 2.1860 to Main Steam Tunnel Unchoked 8 7 8 Main Steam Tunnel to 1st Quadrant199.8 132.4 13.3 - - - - 2.7530 Main Steam Tunnel Unchoked 9 12 9 1st Quadrant Upper Valve Chamber 121.0 16.4 5.7 - - - - 1.5207 to Lower Valve Chamber Unchoked 10 9 10 Openings Between 1st Quadrant 100.0 11.2 4.5 - - - - 1.5120 Upper Valve Chambers Unchoked 11 11 10 1st Quadrant Upper Valve Chamber 121.0 16.4 5.7 - - - - 1.5207 to Lower Valve Chamber Unchoked
- See Figures 1 and 2. f Length/area is the inertial term input directly into RELAP4/MOD5.
Number in parentheses indicates the volume number of the break location which caused choke flow in the vent.
For break locations not indicated, unchoked flow had occurred for the vent.
B/B-UFSAR C3.6-14 REVISION 7 - DECEMBER 1998 TABLE 3 (Cont'd) FROM TO DESCRIPTION VENT VOL. VOL. OF HYDRAULIC HEAD LOSS, K PATH NODE NODE VENT PATH FLOW* AREAf LENGTHf DIAMETER FRICTIONTURNING EXPAN- CONTRAC- NO. NO. NO. CHOKED UNCHOKED ft2 ft ft K, ft/d LOSS, K SION, K TION, K TOTAL 12 12 11 Openings Between 1st Quadrant 100.0 16.1 7.1 - - - - 1.5071 Lower Valve Chambers Unchoked 13 8 13 1st Quadrant Main Steam Tunnel 373.6 42.0 17.6 - - - - 2.2600 Unchoked 14 13 14 1st Quadrant Main Steam Tunnel 270.8 42.0 13.9 - - - - 2.2600 Unchoked 15 8 11 1st Quadrant Lower Valve Chamber 73.0 17.2 4.5 - - - - 1.5685 to Main Steam Tunnel Unchoked 16 8 12 1st Quadrant Lower Valve Chamber 73.0 17.2 4.5 - - - - 1.5685 to Main Steam Tunnel Unchoked 17 2 3 Openings Between 2nd Quadrant 100.0 11.2 4.5 - - - - 1.5120 Upper Valve Chambers Unchoked
- See Figures 1 and 2. f Length/area is the inertial term input directly into RELAP4/MOD5.
Number in parentheses indicates the volume number of the break location which caused choke flow in the vent.
For break locations not indicated, unchoked flow had occurred for the vent.
B/B-UFSAR C3.6-15 REVISION 7 - DECEMBER 1998 TABLE 3 (Cont'd) FROM TO DESCRIPTION VENT VOL. VOL. OF HYDRAULIC HEAD LOSS, K PATH NODE NODE VENT PATH FLOW* AREAf LENGTHf DIAMETER FRICTIONTURNING EXPAN- CONTRAC- NO. NO. NO. CHOKED UNCHOKED ft2 ft ft K, ft/d LOSS, K SION, K TION, K TOTAL 18 2 1 HVAC Panels in 2nd Quadrant 51.3 15.2 5.9 - - - - 2.900 Upper Valve Chambers Choked (5, 6) 19 3 1 Door and HVAC Panels in 2nd 75.8 25.3 5.4 - - - - 2.900 Quadrant Upper Valve Chambers Choked (5, 6) 20 9 1 HVAC Panels in 1st Quadrant Upper Valve Chamber 51.3 15.2 5.9 - - - - 2.900 Choked (5)§ 21 10 1 Door and HVAC Panels in 1st 75.8 25.3 5.4 - - - - 2.900 Quadrant Upper Valve Chamber Choked (5)4 22(a)** 0 5 Main Steam Line Break in Node 5 1.0 0.0 0.0 - - - - 0.000 Fill 22(b)5 0 6 Main Steam Line Break in Node 6 1.0 0.0 0.0 - - - - 0.000 Fill
- See Figures 1 and 2. f Length/area is the inertial term input directly into RELAP4/MOD5.
Number in parentheses indicates the volume number of the break location which caused choke flow in the vent.
For break locations not indicated, unchoked flow had occurred for the vent.
§ Choking results for 2nd quadrant valve room are applied to 1st quadrant valve room.
- Four cases were considered each having a different break location.
B/B-UFSAR C3.6-16 REVISION 7 - DECEMBER 1998 TABLE 3 (Cont'd) FROM TO DESCRIPTION VENT VOL. VOL. OF HYDRAULIC HEAD LOSS, K PATH NODE NODE VENT PATH FLOW* AREAf LENGTHf DIAMETER FRICTIONTURNING EXPAN- CONTRAC- NO. NO. NO. CHOKED UNCHOKED ft2 ft ft K, ft/d LOSS, K SION, K TION, K TOTAL 22(c)** 0 7 Main Steam Line Break in Node 7 1.0 0.0 0.0 - - - - 0.000 Fill 22(d)5 0 8 Main Steam Line Break in Node 8 1.0 0.0 0.0 - - - - 0.000 Fill
- See Figures 1 and 2. f Length/area is the inertial term input directly into RELAP4/MOD5.
Number in parentheses indicates the volume number of the break location which caused choke flow in the vent.
For break locations not indicated, unchoked flow had occurred for the vent.
§ Choking results for 2nd quadrant valve room are applied to 1st quadrant valve room.
- Four cases were considered each having a different break location.
B/B-UFSAR C3.6-17 REVISION 7 - DECEMBER 1998 TABLE 4 UNIT 1 BLOWDOWN RATES AND ENTHALPY FOR MAIN STEAMLINE BREAK TIME (sec) FLOW (lb/sec) ENTHALPY (Btu/lb) 0.0 14,189 1,024.9
0.02 14,189 1,024.9
0.04 13,883 1,116.4
0.06 12,901 1,119.3
0.08 12,479 1,124.6 0.10 12,342 1,133.0 0.12 12,250 1,134.0
0.14 12,093 1,128.3
0.16 11,827 1,119.4
0.18 11,434 1,111.2
0.20 11,022 1,107.3 0.22 10,642 1,105.7
0.24 10,315 1,105.5
0.26 10,041 1,105.9
0.28 9,810 1,106.3
0.30 9,608 1,106.4 0.32 9,424 1,106.4
0.34 9,255 1,106.3
0.36 9,097 1,106.4
0.38 8,954 1,106.8
0.40 8,827 1,107.7 0.42 8,728 1,110.1 0.44 8,633 1,109.1 B/B-UFSAR C3.6-18 REVISION 7 - DECEMBER 1998 TABLE 4 (Cont'd) TIME (sec) FLOW (lb/sec) ENTHALPY (Btu/lb) 0.46 8,558 1,112.7 0.48 8,522 1,116.4
0.50 8,512 1,119.5
0.52 8,523 1,122.7
0.54 8,553 1,125.9
0.56 8,598 1,128.8 0.58 8,652 1,131.3 0.60 8,709 1,133.3
0.62 8,765 1,134.8
0.64 8,813 1,135.8
0.66 8,852 1,136.5
0.68 8,879 1,136.9 0.70 8,894 1,137.2
0.72 8,898 1,137.3
0.74 8,892 1,137.4
0.76 8,878 1,137.5
0.78 8,875 1,138.5 0.80 8,846 1,137.5
0.82 8,816 1,137.9
0.84 8,788 1,138.3
0.86 8,761 1,138.5
0.88 8,733 1,138.7 0.90 8,705 1.138.9 0.92 8,679 1,139.1 B/B-UFSAR C3.6-19 REVISION 7 - DECEMBER 1998 TABLE 4 (Cont'd) TIME (sec) FLOW (lb/sec) ENTHALPY (Btu/lb) 0.94 8,655 1,139.3 0.96 8,632 1,139.5
0.98 8,611 1,139.6
1.00 8,591 1,139.8
1.02 8,573 1,139.9
1.04 8,557 1,140.1 1.06 8,542 1,140.2 1.08 8,529 1,140.3
1.10 8,518 1,140.4
1.12 8,508 1,140.5
1.14 8,499 1,140.5
1.16 8,492 1,140.5 1.18 8,486 1,140.4
1.20 8,481 1,140.3
1.22 8,477 1,140.2
1.24 8,473 1,140.0
1.26 8,471 1,139.7 1.28 8,468 1,139.4
1.30 8,466 1,139.0
1.32 8,465 1,138.6
1.34 8,463 1,138.1
1.36 8,462 1,137.5 1.38 8,462 1,136.9 1.40 8,461 1.136.2
B/B-UFSAR C3.6-20 REVISION 7 - DECEMBER 1998 TABLE 5 UNIT 2 BLOWDOWN RATES AND ENTHALPY FOR MAIN STEAMLINE BREAK TIME (sec) FLOW (lb/sec) ENTHALPY (Btu/lb) 0.0 11,000 1,195.4 2.0 10,434 1,195.1 4.0 9,608 1,196.9 6.0 9,017 1,197.7 8.0 8,613 1,199.4 10.0 9,318 1,199.8 10.1 2,098 1,201.1 20.0 1,993 1,199.2 30.0 1,879 1,208.1 50.0 1,625 1,206.1 75.0 1,064 1,203.0 100.0 814 1,201.5 B/B-UFSAR C3.6-21 REVISION 7 - DECEMBER 1998 TABLE 6 SUMMARY OF UNIT 1 PEAK PRESSURES BETWEEN VALVE ROOM AND MAIN STEAM TUNNEL WALL LOCATION BETWEEN NODES DELTA P ACROSS PEAK PRESSURE (psid) 2-3 9-10 Vertical Wall 5.48 3-4 10-11 Horizontal Floor 12.50 4-5 11-12 Vertical Wall 13.20 2-5 9-12 Horizontal Floor 12.50 4-6 5-6 Main Steam Tunnel 11-8 12-8 Vertical Wall 14.40
B/B-UFSAR REVISION 7 - DECEMBER 1998
ATTACHMENT 3.7A REEVALUATION AND VALIDATION OF THE BYRON/BRAIDWOOD SEISMIC DESIGN BASIS
B/B-UFSAR 3.7A-i ATTACHMENT 3.7A TABLE OF CONTENTS PAGE 3.7A.1 INTRODUCTION 3.7A-1
3.7A.2 BYRON/BRAIDWOOD GENERAL SEISMIC REEVALUATION QUESTIONS/RESPONSES 130.6, 130.6A, 110.68, AND 110.70 3.7A.3 BYRON/BRAIDWOOD ASSESSMENT OF PIPING SYSTEMS, COMPONENTS AND SUPPORTS USING THE NEW SEISMIC RESPONSE SPECTRA - QUESTION/RESPONSE 110.66 3.7A.4 BYRON/BRAIDWOOD DETAILED REVIEW OF THE SEISMIC DESIGN OF ONE PIPING SUBSYSTEM - QUESTION/ RESPONSE 110.67 3.7A.5 BYRON SITE SPECIFIC SEISMIC CONCERNS -QUESTIONS/ RESPONSES 130.9 AND 130.9A 3.7A.6 BRAIDWOOD SITE SPECIFIC SEISMIC CONCERNS - QUESTIONS/RESPONSES 130.56 AND 362.11
B/B-UFSAR 3.7A-1 REVISION 7 - DECEMBER 1998 ATTACHMENT 3.7A 3.7A.1 INTRODUCTION This section describes the Questions and Responses documenting the historical evaluation of the Byron/Braidwood seismic design. It also validates the current seismic design basis which uses the deconvolution analysis to generate response spectra at the foundation level. The Construction Permit for Byron and Braidwood was based on OBE and SSE levels that were 0.09g and 0.20g, respectively, at the foundation. The foundation level spectrum shape was obtained through a deconvolution analysis with the wide band Regulatory Guide 1.60 spectra defined at grade level. Mean soil properties were used in the deconvolution analysis.
Subsequent to the issuance of the Construction Permit and during the review of the FSAR for an Operating License, several aspects of the seismic design were reevaluated. The most notable seismic concern was the reevaluation of the Byron/ Braidwood design using the Regulatory Guide 1.60 spectra without the application of a deconvolution analysis. The Marble Hill design basis, which was a replicate of the Byron/ Braidwood design but had implemented Regulatory Guide 1.60 spectra, was used for comparison in the design assessment.
The reevaluation of the Byron/Braidwood seismic design took place over several years. It included responses to many NRC questions during the review of the FSAR. In order to maintain the chronological perspective of the individual assessments requested by each NRC question, the individual questions and their specific responses have been kept intact in this attachment. The five remaining sections of this attachment have been organized according to topics. NRC questions with Applicant responses for each specific topic have been grouped into the appropriate section. All tables and figures associated with each response have been left intact. A brief description to each question and response is also provided. 3.7A.2 BYRON/BRAIDWOOD GENERAL SEISMIC REEVALUATION During the review of the FSAR, the NRC disallowed the use of the deconvolution analysis to generate the response spectra at the foundation that had been approved during the PSAR review.
Question 130.6 requested a seismic analysis based on Regulatory Guide (R.G.) 1.60 spectra. The response provides a comparison of the Byron/Braidwood design basis and the R.G. 1.60 spectra. It also provides a historical review of the seismic issue and a justification for areas of nonconformance to R.G. 1.60 spectra. Question 130.60a requested a comparison of the structural responses of Category I structures and the design parameters used as the design basis with those which would have been obtained using the R.G. 1.60 response spectra. The response spectra provides a reassessment of the structures in accordance with the question.
B/B-UFSAR 3.7A-2 The Response to Question 130.6a references responses to Questions 110.68 and 110.70. These responses are provided in this section for completeness. QUESTION 130.6 "The seismic analysis was performed by the response spectrum method. However, the response spectra at the foundation level generated by the synthetic time history have displayed a significant dip over a large range of frequencies as compared with the design response spectra in R.G. 1.60 (Figures 3.7-1 through 3.7-40). The use of such unconservative response spectra is unacceptable to the staff. The deconvolution procedure as described in the FSAR is not appropriate for the Byron/Braidwood sites due to the shallow soil overburden (16 ft to 38 ft) on bedrock. Therefore, it is requested that the analysis shall be based on R.G. 1.60 free field surface design response spectra applied at the foundation level and the design time history shall generate response spectra envelope the R.G. 1.60 design response spectra at the foundation level."
RESPONSE 1. Introduction The seismic design process involves various steps. These include (1) determination of g level; (2) specification of the shape of the design response spectra and design time-history; (3) analysis to obtain design response spectra at the base mat elevation; (4) modeling of the structure; (5) calculation of structural response and floor response spectra; (6) specification of load factors, load combinations, factors of safety, and allowable stresses; and (7) designs of components to the combined effects of seismic and other loads. The overall safety of the plant is a function of the design parameters assumed at each stage. The margin in design for the various stages may vary, but good engineering design requires that the overall design be conservative. In this response, the background to the present design bases is presented to highlight the bases of the present design criteria. The conservatism associated with the design "g" levels, with the design response spectrum when the effect of earthquake wave passage is considered, with the use of elastic analysis and low damping values, and the use of minimum yield/ultimate strength for design are quantified to show that any reduction in response due to the use of the deconvolution analysis is more than compensated for by the margins in design introduced by the conservative definitions of other seismic design parameters. To comply with the NRC request, the structural responses B/B-UFSAR 3.7A-3 and floor response spectra obtained by applying the Regulatory Guide 1.60 spectra at the foundation levels are also presented.
Based on a composite evaluation of the above information it is concluded that the present design of Byron/Braidwood is conservative. 2. Background to Present Design Criteria In the present Byron/Braidwood design, a wide band Regulatory Guide 1.60 spectrum is specified at the grade elevation. One dimensional deconvolution analysis is used to compute the foundation elevation spectra. The structural response and the floor response spectra are computed using the deconvolved foundation level spectra. In the PSAR, it was our evaluation that for the Byron and Braidwood sites a 0.06 g OBE and a 0.12 g SSE level are conservative design bases. The NRC staff stated in Question 2.5.63 that the OBE and SSE levels should be 0.1 g and 0.20 g, respectively. In the ensuing discussions with the NRC staff, it was agreed that the OBE and SSE spectra at the foundation elevation will have 0.09 g and 0.20 g rigid period accelerations, respectively. The foundation level spectrum shape was to be obtained through a SHAKE (Reference 1) deconvolution analysis with the wide band Regulatory Guide 1.60 spectra defined at the grade elevation. Consistent with the practice at that time (1974), mean soil properties were used in the deconvolution analysis. The PSAR was amended in November 1974 to reflect the above design bases. In December 1975, the construction permit was issued by the NRC. In September 1976, the NRC requested additional information, stating that: "The current NRC staff position is that when the design response spectra are defined for the free field and applied at the finished grade level of the site, the SHAKE computer program is acceptable for deconvolution analysis to obtain a time history at the base of the idealized soil profile provided that appropriate soil properties, and variations thereof, are used in the analysis.
"In view of the uncertainty and variability of soil properties, the response spectra at the base of the soil-structure interaction system should envelop all response spectra of those deconvolved time histories within the range of variable soil properties, and should not be less than 60 percent of the free field surface spectra."
B/B-UFSAR 3.7A-4 A reply to the above NRC concern was submitted on December 9, 1976 (Reference 2). In the reply it was stated: "We found that a variation in soil properties of +/-20% and a strict adherence to the requirements of Standard Review Plan 3.7.1 that the foundation spectrum be no less than 60% of the surface spectrum at any point would cause an increase in the design forces for Category I structures. Most of the increase in forces is due to the rather arbitrary 60% limit and not due to the +/-20% soil property variation.
"There are several areas of conservatism in the seismic analysis for Byron/Braidwood. Areas such as the methods used for the determination of the maximum ground acceleration for the SSE and the OBE have a considerable amount of conservatism. The use of the wide band response spectrum and the corresponding synthetic time history that envelopes the spectrum is another factor which results in higher forces than actual. Various items in the modeling and analysis, such as lower damping values, three simultaneous spatial components of equal strength, not accounting for the traveling nature of seismic waves are all areas of conservatism which are built into the analysis. "We have also reviewed the conservatism in many of the assumptions and methodology used in the design, compared the actual material strength obtained in the field with the design strength used, and have concluded that the increases in the design forces are more than compensated by these conservatisms. Therefore, the overall safety margin of the stations is not affected. Since our response was accepted by the NRC and no further information was requested, the design and construction of the Byron/Braidwood plant proceeded, based on the design criteria as contained in the PSAR. At the present time, the structural design and construction of the plant structures are complete. The remaining electrical and mechanical components and equipment are either on site or at advanced stages of fabrication and qualification. It is evident from the above that the present Byron/ Braidwood design criteria were appropriately judged to be conservative by the NRC staff in 1974 and again in 1976. It is also clear that the judgment was based on an overall evaluation of the seismic design process. We feel that none of the parameters have changed since then to alter this conclusion.
B/B-UFSAR 3.7A-5 3. Conservative Selection of Design Earthquakes For the selection of design earthquakes, the maximum historical random earthquake of the entire seismotectonic province is assumed to occur at the site even if there is no history of seismic activity in the site vicinity. This is a very conservative assumption. In addition, the staff has required that a VII-VIII intensity earthquake be considered. In the case of the Byron/Braidwood sites, it is our conclusion that the maximum random earthquake should be of MM Intensity VII. Our reasons for this have been documented in detail in the Byron/Braidwood PSAR. Using MM Intensity VII, the Trifunac & Brady relationship gives a maximum acceleration of 0.13 g. A more recent NRC-sponsored study (Reference 3), performed by Computer Services Corporation (CSC), which was based on much more exhaustive data, yields an acceleration of 0.085 g for the United States sites. Even for an Intensity VII-VIII earthquake, the CSC study gives only an 0.11 g level for United States sites. On the basis of the above reasoning, the value of 0.20 g for SSE is higher than necessary, and a value of 0.12 g, as proposed during the PSAR review stage, is more than appropriate.
For the OBE, the design acceleration is 0.09 g. The bases used for this acceleration are extremely conservative when compared with the more recent projects. A seismic risk analysis for the Byron/Braidwood Stations showed that the return period for an Intensity VI earthquake would be 2150 years. This return period is high when compared to the return period used in the Koshkonong (1000 years) project, which is more recent. The return periods for Intensities IV and V at the Byron/Braidwood site would be 322 years and 833 years, respectively. These return periods are more comparable to the Koshkonong project. Thus, a more appropriate OBE intensity for Byron/Braidwood would be IV or at most V. The acceleration values obtained from Trifunac & Brady and the CSC relationships for these Intensities are shown in Table Q130.6-1. It can be concluded from Table Q130.6-1 that 0.06 g is a more reasonable acceleration level for the OBE for the Byron/Braidwood design. The 0.06 g level for OBE was proposed in the initial PSAR submittal. Based on the above discussion, levels of 0.06 g for OBE and 0.12 g for SSE can be considered appropriate but conservative design bases. Figures Q130.6-1 and Q130.6-2 provide a comparison of the Byron/Braidwood deconvolved design spectra to the 0.06 g OBE and 0.12 g SSE Regulatory Guide 1.60 spectra for horizontal and vertical motions, respectively. The comparison shows that the Byron/Braidwood design bases B/B-UFSAR 3.7A-6 envelop the Regulatory Guide spectra. Thus, the seismic forces obtained by applying a 0.06 g OBE and 0.12 g SSE Regulatory Guide 1.60 spectra would be smaller than those presently considered in the design for the containment, containment internal structures, and the auxiliary fuel handling building complex. Table Q130.6-2 shows the comparison for overturning moment and base shear force for the containment shell structure. The total shear force and overturning moment from the regulatory guide input are lower than those used for design. Figures Q130.6-9 through Q130.6-56 provide a comparison of the present design floor response spectra with those obtained using 0.06 g OBE and 0.12 g SSE Regulatory Guide 1.60 spectra. OBE spectra are for 1% oscillator damping, whereas SSE spectra are presented for 2% oscillator damping. The comparison is provided for both horizontal and vertical responses in the containment and the auxiliary building complex. Table Q130.6-3 lists the location and elevations for the spectra comparison. The comparison shows that the present design spectra are higher and, except for a few isolated instances, they envelop those obtained using the Regulatory Guide 1.60 spectra.
From the above discussions, it can be concluded that the Byron/Braidwood design is conservative. Any reduction in response due to deconvolution is more than compensated by the extremely conservative specification of the OBE and SSE levels. 4. Design Spectra Considering Effect of Foundation Size The observation has frequently been made that structures on large foundations appear to respond with less intensity to earthquakes than do smaller structures and, more specifically, than does free field instrumentation. Researchers who have attempted to give a rational explanation for this behavior have concluded that during an earthquake, not all particles under a large building foundation describe the same motion simultaneously; thus the relatively rigid structure-foundation system tends to average the ground motion, resulting in a reduced effective input excitation and consequently less damage. In a report to the NRC dated September 1976 and entitled "A Rationale for Development of Design Spectra for Diablo Canyon Reactor Facility," Dr. Newmark investigated the effect of foundation size on design spectrum (Reference 4). His recommended reduced effective inputs, and the earthquake wave transit times, , for various structures, are given in Table Q130.6-4.
It is recognized that the reduced effective spectra were developed by Dr. Newmark for the Diablo Canyon site and for a near-field earthquake on a rock site. However, it is our B/B-UFSAR 3.7A-7 evaluation that the concept and the methodology proposed by Dr. Newmark are also applicable to the Byron/Braidwood seismic design. Our evaluation is based on a comparison of the seismic design parameters for the two plants. The Byron/Braidwood and Diablo Canyon building sizes and rock site conditions are comparable; wave transit times of 0.04 second for the containment and 0.067 second for the auxiliary-turbine building complex are appropriate. For the Byron/Braidwood site, the maximum historical earthquake of the entire seismotectonic province is assumed to occur at the site. Thus, the earthquake is, by definition, a near-field earthquake and the reduced effective spectra due to wave transit times can be constructed using the reduction factors recommended by Dr. Newmark. It is possible that ground motions at the Byron/Braidwood site may occur due to seismic activity at distances greater than those considered for the reduced effective spectra at the Diablo Canyon plant. However, the ground motions in such an event are likely to be smaller than the design basis ground motions and would not control the design. Figures Q130.6-3 and Q130.6-4 present a comparison of the 0.09 OBE and 0.20 SSE Regulatory Guide 1.60 spectra, the deconvolved Byron/Braidwood design bases spectra, and the reduced effective spectra (denoted as "" spectra on the figures) for the containment and the auxiliary-turbine building complex for OBE and SSE, respectively. The hatched area shows the frequency region where the Byron/Braidwood spectra are exceeded by the reduced effective spectra. It can be observed that for the auxiliary-turbine building complex the Byron/Braidwood design spectra envelop the reduced effective spectra. For the containment building, the Byron/Braidwood spectra do not fully envelop the reduced effective spectra; however, at the predominant structural period of 0.287 seconds, the Byron/Braidwood spectra are higher. Thus, it can be concluded that when the effect of foundation size on the design spectra is considered, the present Byron/Braidwood seismic design is conservative.
- 5. Conservatism in Analysis As in the seismic analysis of any complex structure, several conservative assumptions are used in the Byron/Braidwood design. Many of these assumptions are regulatory requirements; others were necessary to simplify the analysis. These assumptions do provide additional margins of safety. They are briefly described below.
The time-history used for Byron/Braidwood (B/B) deconvolution analysis has a response spectrum which is 5% to 20% higher than the Regulatory Guide 1.60 spectrum in the range of significant structural frequencies (4-12 Hz). This is shown in Figures Q130.6-5 and Q130.6-6 for 4% and 7% damping, respectively.
B/B-UFSAR 3.7A-8 In the B/B design, the two horizontal and the vertical simultaneous components of earthquake motion are assumed to have the same maximum accelerations as required by Regulatory Guide 1.60. However, recorded earthquake motions show that the three components do not have the same accelerations. Studies presented in References 5 and 6 indicate that a 1.0:0.87:0.70 ratio for the three components is more appropriate. Dr. Newmark, in a report (Reference 7) prepared for the NRC, recommends that the vertical acceleration be 2/3 of the horizontal. In the seismic modeling of the containment and the auxiliary/fuel handling/turbine building below the grade level, the effect of the soil or rock on the sides of the exterior walls was neglected in computing the responses. However, the walls were designed for dynamic earth pressures. Consideration of the side soil/rock effect would tend to reduce the overturning moment on the shear walls and the foundation mat. The maximum seismic response of the structure is strongly influenced by the energy absorption characteristics or damping of the structure. Low values of damping result in higher responses and are thus conservative. In the B/B design the damping values recommended in Regulatory Guide 1.61 were used. Newmark and Hall (Reference 8) in their recent report NUREG-0098, prepared for the NRC Systematic Evaluation Program, have recommended higher and more realistic damping values. A comparison of the Regulatory Guide 1.61 damping values and the NUREG-0098 damping values is provided in Table Q130.6-5. Note that the NUREG damping values are higher and thus would lead to lower responses. In considering the response of nuclear power plant structures to seismic motions, one must take into account the implications of various levels of damage short of impairment of the safety, and definitely short of the collapse of the structure. Some elements of plant structures must remain elastic or nearly elastic in order to perform their allocated safety function. However, in many instances, a purely linear elastic analysis may be unreasonably conservative when one considers that even up to the near yieldpoint range there are nonlinearities of amounts sufficient to reduce the required design levels significantly. Moreover, limited yielding of a structure may reduce the response of equipment located in the structure below those levels of response that would be excited were the structure to remain elastic. The concept of ductility factors (Reference 9) is a simple but effective means of accounting for small excursions into the inelastic range. A ductility of 1.3 for concrete and 3.0 for steel members was proposed for the Diablo Canyon Power Plant docket and for the NRC Systematic Evaluation Program Seismic Criteria (Reference B/B-UFSAR 3.7A-9 9). Use of these ductility factors on B/B would result in a 10%-50% reduction in design response computed using an elastic analysis. Based on the above discussion, it can be concluded that there are several areas of major conservatism in the B/B seismic design, and due consideration should be given to these factors when reviewing the B/B seismic design. 6. Conservatism in Material Strength The compressive strength of concrete obtained from the cylinder tests exceeds the value used in design. The actual strength for the reinforcement steel and structural steel also exceeds those used in design. Table Q130.6-6 compares the values used in the design to those obtained by tests. The actual strength is the mean value obtained from the concrete cylinder test report summaries and a sampling of certified material test reports for reinforcement and structural steel for the B/B project. It shows that the actual strength exceeds the design strength by 12% to 50%, adding proportionally to the design margins.
- 7. Response Due to 0.09 g OBE and 0.20 g SSE Regulatory Guide 1.60 Spectrum To comply with the NRC request, forces, moments, and floor response spectra obtained by applying 0.09 g OBE and 0.20 g SSE Regulatory Guide 1.60 spectra at the foundation level are compared to the corresponding B/B design forces, moments, and floor response spectra. The comparison is provided for the containment and the auxiliary building. The forces, in many instances, increase when the Regulatory Guide 1.60 spectra are applied. However, these increased forces should be judged against the conservatism in the B/B design as discussed previously. a. Containment Forces Table Q130.6-7 presents a comparison between forces and moments for the containment shell and base mat between current B/B values and those obtained by applying the Regulatory Guide 1.60 spectra at the base level. The base mat is designed to resist overturning moments from the containment shell and the containment internal structures. The magnitude of these moments affects the area of the base mat which is uplifted (see Figure Q130.6-7) and the design moments in the mat (see Figure Q130.6-8). The increased overturning results in the engagement of the reactor cavity as a rotational key producing large meridional membrane forces, whereas for the current B/B design these forces are negligible. Note that in the seismic modeling of the containment below the grade level, the effect of the soil B/B-UFSAR 3.7A-10 or rock on the sides of external walls was neglected in computing the responses. Consideration of the side/rock effect would tend to reduce the overturning moments and meridional membrane forces. b. Containment Internal Structures 1) Reinforced Concrete A review of the internal concrete structures including the refueling pool walls, primary shield wall, secondary shield wall, and enclosure walls was made. The seismic design is controlled by forces generated from horizontal SSE spectra. The lowest horizontal frequency for the internal structures model is 9.8 cps. Figure Q130.6-81 shows that for 9.8 cps and higher frequencies, the B/B design spectra envelop the Regulatory Guide 1.60 spectra. The resulting forces for the Regulatory Guide 1.60 spectra are consequently lower than those used in the present B/B design. 2) Structural Steel A summary of the structural steel beams showing the percent increase in force for OBE and SSE conditions due to the application of Regulatory Guide 1.60 spectra is presented in Table Q130.6-8. The table is based on a representative sample comprised of all beams at elevation 426 feet 0 inch. A total of 108 beams were reviewed, out of an estimated 740 per unit. Note that the increase in forces in 100 of the 108 beams is less than 20%.
Structural steel columns are seismically designed by amplifying the permanent loads on the columns in proportion to the zero period acceleration of the wall response spectra at that elevation. The minimum values for g used for design are 0.5 and 0.9 for OBE and SSE, respectively. The maximum values for the zero period acceleration of the wall spectra at various elevations for the Regulatory Guide 1.60 foundation elevation definition are 0.26 for OBE and 0.42 for SSE. Comparing these g values shows that the forces are consequently larger for the present B/B design than for the Regulatory Guide 1.60 foundation elevation definition.
B/B-UFSAR 3.7A-11 c. Auxiliary-Fuel Handling Building Complex 1) Reinforced Concrete The areas of the base mat found to have increased forces as a result of Regulatory Guide 1.60 spectra are indicated by the cross-hatched areas shown in Figure Q130.6-8. The values shown are the percent increase in force over the design force.
The comparison of all the shear wall forces from the B/B design basis with those resulting from implementation of the Regulatory Guide 1.60 spectra was made. A summary of this comparison, showing the percent change in seismic force due to Regulatory Guide 1.60 spectra is shown in Table Q130.6-9. Note that for the SSE excitation, the increase in forces in 234 out of 272 shear walls is less than 20%.
The remaining concrete structural components, including columns, beams, and slabs, have no increase in seismic forces due to the implementation of Regulatory Guide 1.60 spectra. These components are essentially rigid and are located at elevation 401 feet 0 inch or below. A comparison of the vertical spectra in Figures Q130.6-59 and Q130.6-83 shows that the B/B spectra envelop the Regulatory Guide 1.60 spectra. 2) Structural Steel A summary of the structural steel beams for the auxiliary-fuel handling building complex showing the percent increase in force due to the application of Regulatory Guide 1.60 spectra is presented in Table Q130.6-10. A representative sample comprised of those beams at elevations 451 feet 0 inch and 426 feet 0 inch has been reviewed. The 230 beams reviewed are representative of all beams. Note that for the SSE excitation the increase in forces in 224 of the 230 beams is less than 20%. The criteria for the design of the structural steel columns in the auxiliary-fuel handling complex are the same as those for the containment building. The minimum value for g used for design is 0.26 for OBE and 0.68 for SSE. The maximum value for various elevations, based on a Regulatory Guide 1.60 spectra foundation elevation definition is 0.23 for OBE and 0.46 B/B-UFSAR 3.7A-12 for SSE. Comparing these "g" values shows that the forces are consequently larger for the present B/B design than for Regulatory Guide 1.60. d. Floor Response Spectra Figures Q130.6-57 through Q130.6-104 provide a comparison of the present design floor response spectra with those obtained by using 0.09 g OBE and 0.20 g SSE Regulatory Guide 1.60 spectra at the foundation elevation. The comparison is provided for both horizontal and vertical response for the containment and auxiliary building. Table Q130.6-11 lists the locations and elevations for the spectra comparison. 8. Summary In the present Byron/Braidwood design, a wide band Regulatory Guide 1.60 spectrum is specified at the grade elevation. One-dimensional deconvolution analysis is used to compute the foundation elevation spectra. The structural response and floor response spectra are computed using the deconvolved foundation level spectra. In the question, the NRC staff states that the deconvolution procedure is not appropriate for the Byron/Braidwood sites and that analysis should be based on a Regulatory Guide 1.60 spectra applied at the foundation elevation. In this response, the background to the present design is presented to highlight the bases of the present criteria. The conservatism associated with the design "g" levels, with the design response spectrum when the effect of earthquake wave passage is considered, with the use of elastic analysis and low damping values, and the use of minimum yield/ultimate strength for design are quantified to show that any reduction in response due to the use of the deconvolution analysis is more than compensated for by the margins in design introduced by the conservative definition of other seismic design parameters. To comply with the NRC request, the structural responses and floor response spectra obtained by applying the Regulatory Guide 1.60 spectra at the foundation levels are also presented.
Based on the composite elevation of the above information, it is concluded that the present design of Byron/Braidwood is conservative.
B/B-UFSAR 3.7A-13 References 1. P. B. Schnabel, J. Lysmer, and H. B. Seed, "SHAKE - A Computer Program for Earthquake Response Analysis of Horizontally-Layered Sites," Earthquake Engineering Research Center, Report No. EERC 72-12, University of California, Berkeley, 1972. 2. Submittal to the NRC dated December 9, 1976, entitled "Byron and Braidwood Stations Units 1&2 - Additional Information on Seismic Design Analysis," NRC Dockets 50-454/455 and 50-456/457.
- 3. J. R. Murphy and L. J. O'Brien, "Analysis of a Worldwide Strong Motion Data Sample to Develop an Improved Correlation Between Peak Acceleration, Seismic Intensity, and Other Physical Parameters," NUREG-0402, report prepared for the NRC by Computer Services Corporation, January 1978. 4. N. M. Newmark, "A Rationale for Development of Design Spectra for Diablo Canyon Reactor Facility," report to the USNRC, September 1976.
- 5. A. S. Bartu, "Discussion on Seismic Design Spectra for Nuclear Power Plants," Journal of the Power Division, ASCE, December 1974. 6. J. Penzien and M. Watabe, "Characteristics of 3-Dimensional Earthquake Ground Motions," Earthquake Engineering and Structural Dynamics, Vol. 3, 1975. 7. W. J. Hall, B. Mohraz, and N. M. Newmark, "Statistical Studies of Vertical and Horizontal Earthquake Spectra," Report NUREG-0003, prepared for USNRC, January 1976.
- 8. N. M. Newmark and W. J. Hall, "Development of Criteria for Seismic Review of Selected Nuclear Power Plant," Report NUREG/CR-0098, prepared for NRC, May 1978. 9. N. M. Newmark and E. Rosenblueth, "Fundamentals of Earthquake Engineering," Prentice-Hall, Inc., 1971. 10. NRC Summary of Meeting held February 4, 1977 to discuss the Diablo Canyon Seismic Design Re-evaluation.
B/B-UFSAR 3.7A-14 TABLE Q130.6-1 ACCELERATION-INTENSITY RELATIONSHIP INTENSITY TRIFUNAC & BRADY CSC VI 0.065 g 0.05 g V 0.0325 g 0.029 g IV 0.0165 g 0.0165 g B/B-UFSAR 3.7A-15 TABLE Q130.6-2 COMPARISON OF CURRENT B/B CONTAINMENT FORCES WITH 0.12 g REGULATORY GUIDE 1.60 SSE FORCE OR MOMENT (SSE) ITEM B/B DESIGN NRC REGULATORY GUIDE 1.60 (0.12 g) Total overturning moment at base of shell 4,540,000ft-k 3,156,000ft-k Total shear at base of shell 26,500k 18,420k
B/B-UFSAR 3.7A-16 TABLE Q130.6-3 LOCATIONS FOR SPECTRA COMPARISON - B/B DESIGN VS. 0.06 g OBE AND 0.12 g SSE BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKE DIRECTION FIGURE NUMBER Auxiliary & Containment 330;374 OBE Horizontal EW;NS Q130.6-9 Auxiliary & Containment 330;374 OBE Vertical Q130.6-10 Auxiliary (wall) 346;364 OBE Vertical Q130.6-11 383;401 Auxiliary (slab) 346;364 OBE Vertical Q130.6-12 383;401 Auxiliary 401 OBE Horizontal EW Q130.6-13 Auxiliary 401 OBE Horizontal NS Q130.6-14 Auxiliary;Turbine; Heater Bay 426 OBE Horizontal NS Q130.6-15 Auxiliary;Turbine; Heater Bay 426 OBE Horizontal EW Q130.6-16 Auxiliary (wall) 426;439 OBE Vertical Q130.6-17 451 Auxiliary (slab) 426;451 OBE Vertical Q130.6-18 439 B/B-UFSAR 3.7A-17 TABLE Q130.6-3 (Cont'd) BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKE DIRECTION FIGURE NUMBER Auxiliary;Turbine Heater Bay 451 OBE Horizontal NS Q130.6-19 Auxiliary;Turbine Heater Bay 451 OBE Horizontal EW Q130.6-20 Auxiliary 477 OBE Horizontal NS Q130.6-21 Auxiliary 477 OBE Horizontal EW Q130.6-22 Auxiliary (slab) 467;477 OBE Vertical Q130.6-23 Auxiliary (wall) 467;477 OBE Vertical Q130.6-24 473;485 Containment 424;436 OBE Vertical Q130.6-25 Containment 424;436 OBE Horizontal NS Q130.6-26 Containment 496 OBE Horizontal NS;EW Q130.6-27 Containment 496 OBE Vertical Q130.6-28 Containment Inner Structure 426 OBE Horizontal NS Q130.6-29 Containment Inner Structure 426 OBE Horizontal EW Q130.6-30 Containment Inner Structure (wall) 412;426 OBE Vertical Q130.6-31
B/B-UFSAR 3.7A-18 TABLE Q130.6-3 (Cont'd) BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKE DIRECTION FIGURE NUMBER Containment Inner Structure (slab) 390;401 412;426 OBE Vertical Q130.6-32 Auxiliary & Containment 330;374 SSE Horizontal Q130.6-33 Auxiliary & Containment 330;374 SSE Vertical Q130.6-34 Auxiliary (wall) 346;383 SSE Vertical Q130.6-35 364;401 Auxiliary (slab) 346;383 SSE Vertical Q130.6-36 364;401 Auxiliary 401 SSE Horizontal NS Q130.6-37 Auxiliary 401 SSE Horizontal EW Q130.6-38 Auxiliary;Turbine Heater Bay 426 SSE Horizontal NS Q130.6-39 Auxiliary;Turbine Heater Bay 426 SSE Horizontal EW Q130.6-40 Auxiliary (wall) 426;439 SSE Vertical Q130.6-41 451 Auxiliary (slab) 426;439 SSE Vertical Q130.6-42 451 B/B-UFSAR 3.7A-19 TABLE Q130.6-3 (Cont'd) BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKE DIRECTION FIGURE NUMBER Auxiliary;Turbine Heater Bay 451 SSE Horizontal NS Q130.6-43 Auxiliary;Turbine Heater Bay 451 SSE Horizontal EW Q130.6-44 Auxiliary (wall) 467;477 SSE Vertical Q130.6-45 473;485 Auxiliary (slab) 467;477 SSE Vertical Q130.6-46 Auxiliary 477 SSE Horizontal NS Q130.6-47 Auxiliary 477 SSE Horizontal EW Q130.6-48 Containment 424;436 SSE Horizontal NS;EW Q130.6-49 Containment (wall) 424;436 SSE Vertical Q130.6-50 Containment 496 SSE Horizontal NS;EW Q130.6-51 Containment (wall) 496 SSE Vertical Q130.6-52 Containment Inner Structure 426 SSE Horizontal NS Q130.6-53 Containment Inner Structure 426 SSE Horizontal EW Q130.6-54 Containment Inner Structure (wall) 412;426 SSE Vertical Q130.6-55 Containment Inner Structure (slab) 390;412 401;426 SSE Vertical Q130.6-56 B/B-UFSAR 3.7A-20 TABLE Q130.6-4 EARTHQUAKE WAVE TRANSIT TIME AND PEAK GROUND ACCELERATION STRUCTURE (sec) PEAK GROUND ACCELERATION (g) REDUCTION FACTOR Small Structures 0.00 0.75 1.00
Containments 0.04 0.60 0.80 Auxiliary Building 0.052 0.55 0.73 Turbine Building 0.067 0.50 0.67
B/B-UFSAR 3.7A-21 TABLE Q130-6.5 COMPARISON OF REGULATORY GUIDE 1.61 AND NUREG-0098 DAMPING VALUES STRUCTURE OR COMPONENT REGULATORY GUIDE 1.61 DAMPING FOR SSE NUREG-0098 DAMPING AT OR JUST BELOW YIELD Piping 2-3 2 to 3 Welded Steel 4 5 to 7 Prestressed Concrete (a) Without complete loss in prestress 5 5 to 7 (b) With no prestress left 7 7 to 10 Reinforced Concrete 7 7 to 10 Bolted Steel Structures 7 10 to 15
B/B-UFSAR 3.7A-22 TABLE Q130.6-6 MATERIAL STRENGTH MATERIAL DESIGN STRENGTH (psi) ACTUAL MEAN STRENGTH (psi) Concrete 5,500 6,935 ('Cf) 3,500 5,265 Reinforcement 60,000 67,000 (fy) Structural Steel 36,000 43,200 (fy) 50,000 56,000
B/B-UFSAR 3.7A-23 TABLE Q130.6-7 COMPARISON OF B/B CONTAINMENT DESIGN FORCES TO THOSE FROM REGULATORY GUIDE 1.60 SPECTRA DESCRIPTION B/B (SSE) REGULATORY GUIDE 1.60 (0.2 g) Total overturning moment 4,540,000ft-k 5,260,000ft-k at base of shell Total shear at base of 26,500k 30,700k shell Net tensile membrane 27 k/ft 72 k/ft force in shell Bending moment in base mat 6,650ft-k/ft 9,513ft-k/ft Net membrane tensile force NA 1,335 k/ft in reactor cavity wall
B/B-UFSAR 3.7A-24 TABLE Q130.6-8 CONTAINMENT BUILDING STRUCTURAL BEAMS COMPARISON OF FORCES BETWEEN B/B DESIGN BASIS FORCE AND REGULATORY GUIDE 1.60 NUMBER OF BEAMS % CHANGE IN DESIGN BASIS FORCE OBE SSE < 0 84 88 0 - 10 12 8 10 - 20 4 8 20 - 4 30 - 40 8 -- > 40 -- -- TOTAL108 108 NOTE
- 1. All 108 beams reviewed for elevation 426 feet 0 inch.
- 2. % increase does not necessarily reflect a state of stress in the beam.
B/B-UFSAR 3.7A-25 TABLE Q130.6-9 AUXILIARY BUILDING-FUEL HANDLING BUILDING COMPLEX SHEAR WALLS - COMPARISON BETWEEN B/B DESIGN BASIS SEISMIC FORCES AND REGULATORY GUIDE 1.60 NUMBER OF SPRINGS % CHANGE IN DESIGN BASIS FORCES OBE SSE < 0 122 153 0 - 10 25 51 10 - 20 19 30 20 - 30 8 7 30 - 40 50 8 40 - 50 33 6 > 50 15 17 TOTAL272 272 NOTE Percent increase does not necessarily reflect a state of stress in the wall.
B/B-UFSAR 3.7A-26 TABLE Q130.6-10 AUXILIARY-FUEL HANDLING BUILDING COMPLEX STRUCTURAL STEEL BEAMS - COMPARISON OF FORCES BETWEEN B/B DESIGN BASIS FORCES AND REGULATORY GUIDE 1.60 NUMBER OF BEAMS % CHANGE IN DESIGN BASIS FORCES OBE SSE < 0 148 132 0 - 10 61 85 10 - 20 16 7 20 - 30 5 6 > 30 --- --- TOTAL230 230 NOTE 1. Beams located at elevation 426 feet 0 inch and 451 feet and 0 inch in the auxiliary building. 2. Percent increase does not necessarily reflect a state of stress in the beam.
B/B-UFSAR 3.7A-27TABLE Q130.6-11 LOCATIONS FOR SPECTRA COMPARISON - B/B DESIGN VS. 0.09 g OBE and 0.20 g SSE BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKEDIRECTION FIGURE NUMBER B/B-UFSAR 3.7A-28 REVISION 9 - DECEMBER 2002 TABLE Q130.6-11 (Cont'd) BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKEDIRECTIONFIGURE NUMBER B/B-UFSAR 3.7A-29 REVISION 9 - DECEMBER 2002 TABLE Q130.6-11 (Cont'd) BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKEDIRECTIONFIGURE NUMBER B/B-UFSAR 3.7A-30 REVISION 9 - DECEMBER 2002 TABLE Q130.6-11 (Cont'd) BUILDING ELEVATION BYRON/BRAIDWOOD EARTHQUAKEDIRECTIONFIGURE NUMBER B/B-UFSAR 3.7A-31QUESTION 130.6a* "We have reviewed your response to Question 130.6 and we conclude that it is not adequate and not acceptable for the following reasons: 1)Selection of SSE and OBE Design EarthquakesA considerable portion of our response is based on the conservatism you feel is in the SSE and OBE design earthquakes. You also presented arguments for reducing the design earthquake to those originally proposed in the PSAR (zero period acceleration of 0.06g for OBE and 0.12 for SSE).
These values have been subsequently increased to 0.09g and 0.20g respectively and rationale for the Regulatory staff position was stated in the Question 2.5.63. Furthermore, on the basis of further investigation, the staff came to the conclusion that the deconvolution procedures are not acceptable and that the Regulatory Guide 1.60 Design Response Spectra should be applied at the foundation level. 2)Effect of Foundation Size on Design SpectraThe response suggests that the design spectra can be reduced based on previous studies performed by Dr. Newmark for the Diablo Canyon Site. These studies justify reduced effective spectra as a result of considering the effect of foundation size on design spectrum. You pointed out in the response that the reduced effective spectra were developed for the specific site of the Diablo Canyon Plant and the basic reason for its acceptance was the postulated near-field earthquake. Since the Byron/Braidwood sites are located in an entirely different tectonic province, the argument which was used in the case of Diablo Canyon application cannot be applied to the subject sites. 3)Conservatism in AnalysisThe staff does agree that three components of earthquake motion are probably not the same acceleration. The magnitude of the actual acceleration of each component should be found by means of a 3-dimensional analysis. It is the position of the staff that the response spectrum for vertical motion can be taken as 2/3 of the response spectrum for horizontal motion for the Western United States only. For other locations, the vertical response spectrum should be the same as that given in Regulatory Guide 1.60 (see enclosure). *QUESTION 130.6a is a restated version of NRC QUESTION 130.6.
B/B-UFSAR 3.7A-32As far as the damping values are concerned, the referenced report, NUREG-CR 0098 was developed for a specific purpose of evaluating seismic risk of nuclear plants which are already operating. The damping values contained in that report cannot be applied in licensing of new plants. The response claims that the elastic analysis which is used in design of new plants may be unreasonably conservative. In view of the fact that there is a lot of safety-related equipment which might produce catastrophic consequences in case of excessive deformation of supporting members, this position of the Regulatory staff is not unreasonable. You neglected to mention in your response that the referenced criteria for the Diablo Canyon plant stipulate that the ductility of 1.3 for concrete and 3 for steel are for turbine building and intake structure. These structures are non-Category I per se and the only reason that they have been reviewed by the staff was that in certain locations they are housing some safety-related equipment. Thus the criteria which are applicable to these two structures cannot be automatically applied to all Category I structures. Evaluation of Structures using 0.09g OBE and 0.20g SSE Regulatory Guide 1.60 Spectrum The evaluation of structures using the Regulatory criteria provided in the response have been reviewed. It is recognized that there is a general increase in the stress level of many structural members. We find, however, that without re-analysis of the affected structures and determination of the shear forces and moments imposed by the new loads the evaluation cannot be considered to be conclusive. You are, therefore, requested to compare the structural responses of Category I structures and the design parameters (bending moments, shears and axial loads) actually used in design of Byron/Braidwood plant with those which would have been obtained if the criteria stated in Question 130.06 were used." RESPONSE Introduction The design of the Byron/Braidwood structures and components required for safe shutdown was reassessed during Regulatory Guide 1.60 input. This comprehensive review was made for the SSE condition (0.20 ZPA) to ensure adequate safety. The reassessment basis was agreed upon in the meeting at Bethesda on February 26, 1981 with the NRC, Commonwealth Edison Company, and Sargent & Lundy. The Marble Hill design response spectra was utilized to determine the effects on Byron/Braidwood of the Regulatory Guide 1.60 seismic event. Where structural elements B/B-UFSAR 3.7A-33 were identified as unique on Marble Hill, they were reassessed for Byron/Braidwood using the actual material strengths. The Marble Hill design is based on Regulatory Guide 1.60 spectra for an SSE event of 0.20g. The Marble Hill structures are a replicate of the Byron/Braidwood structures. Structural element properties as shown in Figures Q130.6a-1 and Q130.6a-2 are the same for Byron/Braidwood and Marble Hill. For these reasons, the Marble Hill design forces and spectra were used as the reassessment basis for Byron/Braidwood. Although there are a few unique structural features on Marble Hill, the effect of these features is negligible for purposes of seismic analysis.
Comparison of the Byron/Braidwood design basis model with the Marble Hill design basis model indicates that the overall geometry mass, stiffness, and dynamic characteristics are equivalent. The best measure of model equivalence is shown in the comparison spectra given in attached Addendum. Response spectra generated using the Byron/Braidwood design basis model with the same analysis parameters used in the Marble Hill analysis for comparison to the Marble Hill design basis response spectra is also provided in the attached Addendum.
The containment building seismic models used for the Marble Hill are identical to the models used for the Byron/Braidwood Stations.
Reassessment of Structures The Byron/Braidwood structures were reviewed against the Marble Hill structures by comparing the respective design drawings. All unique elements due to the seismic design were identified.
The unique Byron/Braidwood elements were reviewed for the Marble Hill SSE loading conditions using average actual material strengths. For average actual material strengths, the values given in Table Q130.6a-1 were used.
The actual concrete strengths were obtained from the concrete cylinder test results reported by the onsite independent testing agency. The actual steel strengths were obtained from the certified material test results submitted by the material suppliers.
Containment Building Areas where Marble Hill had unique features were identified and a comparison of the forces were tabulated for these areas. These areas include the base mat reinforcing, the vertical post-tensioning tendons, and the reinforcing steel at the main steam penetrations.
B/B-UFSAR 3.7A-34 The Marble Hill and Byron/Braidwood design basis was based on conservative assumptions as is appropriate for initial design purposes. Conservatisms were identified in the design basis analysis and the analysis was refined accordingly. The results of the assessment based on reanalysis show that stresses are below the design basis allowables. Refinement to the analysis include the following: 1. The overall containment overturning moment, axial force and shear used in the assessment were obtained using a shell analysis rather than the beam analysis.
- 2. The stiffness of the elements used to represent the reactor cavity wall was refined to account for initial concrete cracking due to shrinkage, thermal effects and restraint. Containment Internal Structures Concrete Review of the containment internal concrete structures reveals no unique concrete elements. Therefore, the stresses in the containment internal concrete structures are within design basis allowables. Structural Steel Columns Review of the containment structural steel columns reveal no unique column sections. Therefore, the stresses are within the design basis allowables. Structural Steel Beams Eighty-three beams of 740 total beams per unit have unique design due to SSE forces. Reassessing these beams using the average actual material strength indicates a stress level below yield strength.
Auxiliary/Fuel Handling Building Base Mat The auxiliary/fuel handling building base mat contains 17 unique finite elements which represent less than 1% of the total base mat area. (Refer to Figure Q130.6a-3 for unique areas.) Reassessing these elements using the actual average material strength for the concrete and reinforcing steel indicates a stress level below yield strength.
B/B-UFSAR 3.7A-35 Shear Walls Assessment of all shear walls based on Regulatory Guide 1.60 SSE spectra generated loads revealed 27 of the 65 total shear walls in the auxiliary/fuel handling building to have an increase in SSE force. Using the average actual material strengths for the reinforcing steel, the vertical reinforcement for all walls maintains a stress level less than the yield strength. The horizontal reinforcing steel also maintained a stress level less than the yield strength for all walls except for one. This single case has a stress level of 3.5% above the yield strength.
Concrete Beams, Slabs, and Piers There are no unique concrete beams, slabs, and piers between Byron/Braidwood and Marble Hill. Therefore, all stresses are within the design basis allowables. Structural Steel Columns Of the 100 structural steel columns in the auxiliary/fuel handling building, there are 46 columns with at least one unique section. Using the average actual material strength, the stress level did not exceed the yield strength.
Structural Steel Beams A review of the 3,400 structural steel beams in the auxiliary/ fuel handling building revealed 242 unique beams. Reassessing these beams using the average material strengths indicates a stress level below yield strength.
Essential Service Cooling Tower (Byron)
The essential service cooling tower is unique to Byron Station. The SSE forces were generated using Regulatory Guide 1.60 input. Reassessment using these forces indicates that stress levels do not exceed the design basis allowables. Lake Screen House (Braidwood)
The lake screen house is unique to the Braidwood Station. Seismic forces were generated using Regulatory Guide 1.60 input. These forces were lower than the original SSE forces used in the design and, therefore, all stresses remain within the design basis allowables.
Electrical Raceways and Supports The assessment of raceways indicated no unique design features when compared with Marble Hill.
B/B-UFSAR 3.7A-36 Raceway supports on four different elevations in the auxiliary building were chosen for reassessment. These elevations have the largest increase in acceleration values in the frequency range of the raceway support. One hundred and twenty-two hangers, representing 15 different types, were chosen and analyzed using Marble Hill response spectra. The results of reanalysis are given in Table Q130.6a-2. The six hangers with a ductility ratio (actual stress/yield stress) greater than one have overstresses in one of their members giving a ductility ratio ranging from 1.04 to 1.33 with an average of 1.17. These members will not collapse under an SSE event.
Conduit Supports Six conduit supports are unique to Marble Hill. These supports were reassessed and it was determined that Byron/Braidwood supports are within the design basis allowables. HVAC Ducts and Supports The assessment of HVAC ducts indicate no unique design features when compared with Marble Hill.
HVAC duct supports at elevation 477 feet 0 inch in the auxiliary building were reassessed using the Marble Hill spectra. This elevation was chosen because it has the largest increase in acceleration values in the frequency range of the HVAC duct supports. Based on a refined analysis of 347 supports, it was found that stresses do not exceed the yield strength, except for twelve cases. These twelve supports will not collapse under an SSE event. Piping Systems, Supports and In-line Valves Refer to the responses to Questions 110.65, 110.66, 110.67, 110.68, and 110.70. These questions comprise the reassessment for the piping systems, supports, and in-line valves.
Equipment Due to the different methodologies used in reassessing the safe shutdown equipment, the equipment was divided into two groups: NSSS equipment and balance of plant equipment. NSSS Equipment Introduction A reassessment was performed to determine if sufficient margin exists in the NSSS supplied mechanical equipment to withstand a change in the Byron seismic design basis. The investigation was performed for both primary mechanical equipment and
B/B-UFSAR 3.7A-37 auxiliary mechanical equipment. A comparison was made of the generic seismic qualification levels for the equipment with the applicable Marble Hill response spectra. These comparisons indicate that all of the equipment has sufficient design margin to be qualified with no modification to the Marble Hill response spectra. Generic Seismic Analysis of Mechanical Equipment The NSSS supplied mechanical equipment that was included in this study is listed in Table Q130.6a-3. This list contains both primary and auxiliary equipment. All of this equipment was designed and analyzed for use in many plants. As part of the original design, seismic qualification loads were established that provide sufficient margin so that the equipment can be used in plants with both low and high seismic designs. The amount of margin in the design depends on the type of analysis used to qualify the component. Margin for Primary Mechanical Equipment Comparisons were made of the generic seismic response spectra with the Marble Hill response spectra for all of the primary equipment identified in Table Q130.6a-3. An example of the comparison for the reactor vessel internals is shown in Figure Q110.70-1. This figure is typical of all the primary components comparisons and represents the component with the smallest amount of margin between the generic response spectrum and the actual plant response spectrum. Based on these comparisons, it is clear that the response spectra in this reassessment does not significantly change the margin in the components generic design. Margin for Auxiliary Mechanical Equipment All of the auxiliary mechanical equipment was qualified using "g" levels. For the equipment listed in Table Q130.6a-3, the pumps were qualified for 2.1g along each of the two principal horizontal axes and 2g vertically. The tanks and heat exchangers were qualified for 1.5g in each of the two principal horizontal axes and 1.5g in the vertical axes.
All of the pumps included in this study were demonstrated to be rigid by either analysis or test and an investigation of the seismic accelerations for Marble Hill spectra above 33 Hz shows the highest acceleration to be 0.9g. This is well below the 2.1g acceleration used for qualification of the pumps and, therefore, all the pumps are acceptable.
The tanks and heat exchangers all have one modal frequency below 33 Hz. A comparison was made between the qualification levels for these components and the Marble Hill seismic levels. The comparison was made at the calculated component natural frequency with the applicable spectra for the location
B/B-UFSAR 3.7A-38 REVISION 1 - DECEMBER 1989 of the component in the plant. The highest seismic acceleration spectra was found to be 1.1g; this is within the design acceleration for the component of 1.5g. As a result of these comparisons, it was demonstrated that all of the auxiliary mechanical equipment have sufficient margin. Summary A reassessment of the seismic design margin of the NSSS supplied mechanical equipment was performed. The investigation compared the components seismic design levels to the Marble Hill levels. The reassessment demonstrated that all of the components have sufficient design margin.
BOP Equipment The method used to reassess the BOP safe shutdown equipment is as follows: 1. Compare the seismic accelerations, from the appropriate Marble Hill spectra utilizing Regulatory Guide 1.61 damping values against the seismic accelerations used in the existing qualification report. In some cases, the seismic accelerations in the qualification reports exceeded the Marble Hill accelerations. This result can be attributed to one or more of the following: a. Byron spectra exceeding the Marble Hill spectra for the equipment frequency.
- b. The vendor performed a generic qualification.
- c. Vendor enveloped several spectra for equipment located on various elevations. 2. In cases where the Marble Hill levels exceeded the levels in the qualification report, stresses, deflections and margins were calculated for the Marble Hill accelerations by scaling the calculated seismic stresses and deflections by the required Marble Hill to Byron acceleration ratio. 3. Reference response to Question 110.68 for the effect of the Marble Hill spectra on safety-related small piping and instrumentation. Summary Table Q130.6a-4 contains the results. Regulatory Guide 1.61 damping values were used in the reanalysis reassessment. In addition, it should be noted that the resultant stresses were compared to the code allowable stress, not against the failure stress of the material. The results show the safe shutdown equipment can safely withstand the required seismic event.
B/B-UFSAR 3.7A-39RESPONSE TO NRC ENCLOSURE I "INFORMATION REQUIRED FOR REEVALUATION" 1.Item: It is requested that the applicant present theinformation on the respective stress components due to LOCA, if applicable, and SSE and the relation of each with the specified allowable values. Response: The LOCA load condition is only applicable to the containment structures. Reassessment of the containment structure has shown that the stresses are within the design basis allowables. This reassessment included the combined effect of the LOCA and SSE load conditions. 2.Item: Document the following:a.The use of Marble Hill plant seismic analysisas the basis for comparison with the Byron/Braidwood plant. In your discussion, provide acomparison of the mathematical models for keystructures for the two plants which showdynamic parameters such as stiffness, periods,moduli of elasticity, Poisson's ratio andmasses.b.Describe any difference between the two plantsin terms of construction materials, qualitycontrol, construction techniques, etc.c.Describe the detailed procedures as to how theMarble Hill seismic responses will be used incomputing stress levels for different loadcombinations of Byron/Braidwood structuralmembers. In addition, describe the criteriafor selection of members to be evaluated.Response: a. The Marble Hill design is based on RegulatoryGuide 1.60 spectra for an SSE event of 0.20g.The Marble Hill structures are a replicate ofthe Byron/Braidwood structures. For thesereasons, the Marble Hill design forces andspectra were used as the reassessment basis forByron/Braidwood. Although there are a fewunique structural features on Marble Hill, theeffect of these features is negligible formodeling purposes.Comparison of the Byron/Braidwood design basismodel with the Marble Hill design basis modelindicates that the overall geometry, mass,stiffness, and dynamic characteristics are B/B-UFSAR 3.7A-40 equivalent. The best measure of model equivalence is documented in the comparison spectra shown in attached Addendum. Response spectra were generated using the Byron/Braidwood design basis model with the same analysis parameters used in the Marble Hill analysis for comparison to the Marble Hill design response spectra. For completeness, the Byron/Braidwood design basis spectra is also provided in attached Addendum. The Containment Building seismic models used for Marble Hill are identical to the models used for Byron/Braidwood Stations. b. Equivalent material and construction specifications are used on the two plants.
- c. The Byron/Braidwood and Marble Hill structures were compared using their respective design drawings. All unique elements due to seismic design were identified and reassessed using Marble Hill forces and Byron/Braidwood actual material strengths. All design basis loading combinations are the same for Byron/Braidwood and Marble Hill. 3. Item: Provide a comparison of the floor response spectra used in design of structures at key locations for each of the safety-related structures of the Byron/Braidwood and Marble Hill plants.
Response: A comparison of the SSE response spectra is provided in attached Addendum. 4. Item: In order to assess the actual margins in the design for the OBE loads, compare for the key member and the stress levels resulting from the original Byron/Braidwood seismic analysis with those resulting from the use of Marble Hill loads. Response: Reassessment based on OBE loads was done on a few randomly selected structural elements which were reassessed earlier for increased SSE loads. Marble Hill used a zero period acceleration of 0.08g OBE, and Byron/Braidwood used 0.09g OBE; therefore, the Marble Hill loads were factored by 0.09/0.08 to determine the Byron/Braidwood OBE loads.
B/B-UFSAR 3.7A-41 Containment Building Review of the containment structure for Marble Hill loads factored by the ratio between Byron/Braidwood and Marble Hill OBE level indicates that the containment is within design allowable stresses.
Eight structural beams in the containment building were reassessed. Using the average actual material strength, all the beam stresses are within the design basis allowables. Auxiliary/Fuel Handling Building Eleven structural steel beams were selected from the auxiliary/fuel handling building. Using the average actual material strength, a factor of safety between 1.42 to 1.61 was maintained against yield strength.
Ten structural steel columns were reassessed for the auxiliary/fuel handling building. Using the average actual material strength, a factor of safety between 2.43 to 3.35 was maintained against yield strength. Twenty-one shear walls in the auxiliary/fuel handling building were reassessed. Based on the average actual material strength of the reinforcing steel, a factor of safety for the horizontal reinforcing steel between 1.47 to 3.45 was maintained against yield strengths. A factor of safety in the vertical reinforcing steel between 1.61 to 6.58 was also maintained. The stresses in these walls do not exceed the yield strength. From the finite element model of the auxiliary/fuel handling building base mat, ten elements were chosen in the critical SSE area. Using the average actual material strengths for both concrete and reinforcing steel, the stresses of all ten elements did not exceed the yield strength. 5. Item: Ductility under normal conditions shall be one. Exceptions will be considered in special situations for SSE load and where modifications are considered impractical by the Applicant, and the Applicant's judgment is confirmed by the staff, provided that safety is assured. Floor response spectra will be computed on the basis of elastic analysis. Response: Ductility is the ratio between the actual stress to the yield stress. Throughout the reassessment of the structure, a ductility ratio less than or equal B/B-UFSAR 3.7A-42 to one has been maintained. One isolated case for a shear wall where the ductility ratio exceeded one has been indicated. This case is the result of the SSE load condition. Under the OBE condition, the ductility ratio does not exceed one. Seismic responses were completed on the basis of elastic analysis.
- 6. Item: Material Properties a. Concrete As-built compressive strength of concrete, f'c may be used in the reevaluation. The as-built strength of concrete shall be demonstrated by the Applicant through submittal of test data. The average compressive strength, established by the tests, can be used as the "as-built concrete strength." The scope and the extent of tests performed shall be evaluated and approved by the staff. b. Steel Both reinforcing and structural steel yield stresses, fy, will be taken as the average of actual test values. In no case will the yield strength value used in strength computations be taken as greater than 70% of the corresponding tested average ultimate strength value. The scope and the extent of the test program and the resulting test data shall be reviewed and approved by the staff. Response: The average actual material strengths used have been given in Table Q130.6a-1. The actual concrete strength was obtained from the concrete cylinder test results reported by the onsite independent testing agency. The actual steel strengths were obtained from the material certification reports submitted by the steel supplier. The average actual yield strengths have been compared to the average actual ultimate strengths and it was found that the yield does not exceed 70% of the ultimate. 7. Item: Analysis Procedures a. Regulatory Guide 1.61 damping values will be used.
B/B-UFSAR 3.7A-43 b. Accidental torsion will be considered by including an additional eccentricity in the mathematical models of 5% of the building dimension in the direction perpendicular to the applied loads.
- c. Stability requirements as stated in the Standard Review Plan Section 3.8.5 must be met. Response: a. Damping values used in the reassessment conform to Regulatory Guide 1.61. b. The horizontal seismic model for the auxiliary/ fuel handling/turbine buildings include eccentricities corresponding to the distribution of mass and stiffness for Byron/Braidwood. The eccentricities between the mass centroid for a slab and the center of rigidity of its supporting shear walls result in an average torsional moment of 8% of the maximum building dimension times the story shear, for the major slabs in the model. It is unlikely that these eccentricities would increase significantly due to any changes other than major changes to the plant structure since the weight of the permanent structural elements accounts for a large percentage of the total mass. c. Stability requirements of the Standard Review Plan Section 3.8.5 have been met.
Summary The design of the Byron/Braidwood structures and components required for safe shutdown has been reassessed for SSE loads based on Regulatory Guide 1.60. This reassessment has shown that the design of the Byron/Braidwood plant is conservative. Byron/Braidwood seismic design basis ensures that the integrity and functionality of the safety-related structures are maintained.
B/B-UFSAR 3.7A-44 TABLE Q130.6a-1 MATERIAL STRENGTH MATERIAL DESIGN STRENGTH (psi) AVERAGE ACTUAL STRENGTH (psi) Concrete 3,500 5,265 5,500 6,935 Reinforcing Steel (fy) 60,000 67,000* Structural Steel (fy) 36,000 43,200* 50,000 56,000*
- Value does not exceed 70% of the actual average ultimate strength.
B/B-UFSAR 3.7A-45 TABLE Q130.6a-2 RACEWAYS SUPPORTS SUMMARY ELEVATION NUMBER OF HANGERS REVIEWED NUMBER OF SUPPORTS WITH DUCTILITY RATIO 1.0 NUMBER OF SUPPORTS WITH DUCTILITY RATIO >1.0 477 ft-0 in. 55 49 6 451 ft-0 in. 30 30 0 401 ft-0 in. 10 10 0 426 ft-0 in. 27 27 0
B/B-UFSAR 3.7A-46TABLE Q130.6a-3 NSSS MECHANICAL EQUIPMENT I. Primary Mechanical Equipment Reactor Pressure Vessel Reactor Vessel Internals Reactor Coolant Pump Steam Generator Loop Pressurizer II.Auxiliary Mechanical EquipmentA. Pumps Centrifugal Charging Pumps RHR Pumps Boric Acid Pumps Component Cooling Pump Lube Oil Pumps for Centrifugal Charging Pump Essential Service Water Pump Motor Auxiliary Feedwater Pump Motor B. Tanks and Heat Exchangers Boric Acid Tank RHR Heat Exchanger Component Cooling Heat Exchanger B/B-UFSAR 3.7A-47TABLE Q130.6a-4 BOP EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-48TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-49 REVISION 2 - DECEMBER 1990 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-50 REVISION 5 - DECEMBER 1994 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-51 REVISION 1 - DECEMBER 1989 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-52 REVISION 2 - DECEMBER 1990 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-53 REVISION 1 - DECEMBER 1989 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-54TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-55TABLE Q130.6a-4 (Cont'd) DIESEL EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-56TABLE Q130.6a-4 (Cont'd) DIESEL EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-57 REVISION 10 - DECEMBER 2004 TABLE Q130.6a-4 (Cont'd) DIESEL EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-58TABLE Q130.6a-4 (Cont'd) DIESEL EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-59TABLE Q130.6a-4 (Cont'd) DIESEL EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-60TABLE Q130.6a-4 (Cont'd) DIESEL EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-61TABLE Q130.6a-4 (Cont'd) DIESEL EQUIPMENT DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-62 REVISION 1 - DECEMBER 1989 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-63 REVISION 2 - DECEMBER 1990 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-64 REVISION 1 - DECEMBER 1989 TABLE Q130.6a-4 (Cont'd) DESIGN BASIS SPECTRA MARBLE HILL SPECTRA EQUIPMENTLOCATIONH1H2 VERT DAMPING FACTOR (%) NATURAL FREQ (Hz) H1H2 VERT MH SPECTRA ENVELOPED? Y/N IF NO, GIVE TOTAL MARGIN B/B-UFSAR 3.7A-65QUESTION 110.68 "Provide a discussion on the effect of the new seismic response spectra on the safety-related small piping and instrumentation design. Specifically address how the seismic support spans as specified in the Byron/Braidwood 'Small Piping Design Standard' are affected by the new seismic response spectra." RESPONSE All the instrumentation required for safe shutdown was qualified to a spectra enveloping both the design spectra and the Marble Hill spectra. Hence, the new spectra had no effect. In the small piping design procedure, the support spans were governed by the piping stress allowables. In tabulating the support reactions by dynamic analysis, a spectra enveloping both the design and Marble Hill spectra was used. Hence, the new spectra had no effect.
B/B-UFSAR 3.7A-66 QUESTION 110.70 "Indicate an assessment of the effects of the new seismic response spectra on Reactor Pressure Vessel Internals." RESPONSE Figure Q110.70-1 shows the comparison of the generic response spectra to which the reactor pressure vessel internals were qualified versus the design basis and Marble Hill spectra. The generic response spectra is much greater than the Marble Hill spectra. Hence, the Marble Hill spectra has no effect on the reactor pressure vessel internals.
B/B-UFSAR 3.7A-67 3.7A.3 Byron/Braidwood Assessment of Piping Systems, Components, and Supports Using the New Seismic Response Spectra NRC Question 110.66 requested an assessment of the various safety-related piping, components, and supports to meet the new seismic response criteria per Regulatory Guide 1.60. The Applicant used the Marble Hill spectra for the evaluation. The Marble Hill design is a replicate of the Byron/Braidwood design. The Applicant provided a justification for any site unique structural differences at critical sections of each plant.
QUESTION 110.66 "Provide an assessment of those piping systems, components, and supports identified above for the new seismic response spectra. If the applicant chooses to use the Marble Hill spectra for its evaluation of piping and supports, its use must be justified with respect to the structural differences at critical sections of the Byron/Braidwood and Marble Hill plants. a. For piping, provide a table showing the selected critical locations, the design basis seismic stress, the new seismic stress, the total stress (pressure, weight, and seismic), and the allowable stress. b. For valves, provide a similar table showing the acceleration values for the design basis, the new seismic design, and the allowable accelerations.
- c. For supports, provide a similar table showing the support loads for the design basis, the new seismic design, and the support allowable load.
- d. Provide a qualitative measure of the overall margin to failure for each of the above components."
RESPONSE As the Marble Hill design basis is replicate to the Byron and Braidwood design and had implemented Regulatory Guide 1.60 spectra, it was used in the assessment of the Byron/Braidwood piping systems, components, and supports. The structural differences at critical sections of the Byron/Braidwood and Marble Hill plants are as follows.
Methodology Based on the Marble Hill design, the stress level in structural elements unique to Marble Hill as compared to Byron/Braidwood were identified. The overstressed elements were reviewed against the actual average material strength. Reassessed B/B-UFSAR 3.7A-68 elements having a safe shutdown earthquake (SSE) stress level less than the yield limit of the material will be considered acceptable due to the localized nature of the forces. Elements still overstressed will be reviewed on a case by case basis in order to determine their impact on the safe shutdown of the plant. Reassessment of Safety-Related Structures Containment Building Structure The initial assessment of the containment structure was based on a comparison of the Byron/Braidwood design with that of Marble Hill. Areas where Marble Hill had unique design features were identified and force comparisons tabulated. The Marble Hill and Byron/Braidwood design bases, however, were based on conservative assumptions as is appropriate for initial design purposes and, therefore, the force comparison does not reflect the adequacy of the structure. Subsequent to the initial assessment, the following conservatisms were identified and removed and the structure reanalyzed: a. The overall containment overturning moment, axial force and shear used in the reassessment were obtained using a shell model rather than the beam model previously used.
- b. The stiffness of the elements used to represent the reactor cavity wall was modified to account for initial concrete cracking due to shrinkage, thermal effects, and restraints.
- c. Full hydrostatic uplift forces included in the design basis base mat analysis were excluded from the present analysis. This is a more realistic assumption due to the uplift of the base mat during a seismic event. d. The factor applied to the effect of a single horizontal excitation for base mat analysis to account for the effect of three components of excitation was reduced to 1.05 from the 1.10 used in the design basis analysis. The factor 1.10 is based on a study performed during the design basis analysis and it included a 5% to 8% safety margin. The results of the containment structure assessment based on the refined analysis show no overstress in the containment structure.
B/B-UFSAR 3.7A-69Containment Internal Structures Concrete No unique design features were located. Steel Columns No unique design features were located. Steel Beams Eighty-three beams of approximately 740 per unit have a unique design due to seismic forces. These beams were assessed against the Marble Hill SSE response spectra and the results tabulated as follows. Acceptance CriteriaNumber ofBeams Lesser of 0.95 Fy or 1.6 AISC Allowable where Fy = 36 ksi 75 Lesser of 0.95 Fy or 1.6 AISC Allowable where Fy = 42.3* ksi 6 1.0 Fy where Fy = 42.3* ksi 2 83*Actual average material strength for A36 structural steel.Auxiliary/Fuel Handling Building Base Mat The results of the assessment are tabulated as follows. Acceptance Criteria Number of Mat Areas (FiniteElements)ACI Allowable 'cF= 3,500; Fy = 60,000 10 ACI Allowable 'cF= 5,265*; Fy = 67,000* 7 17*Actual average material strength for concrete and Rebar.
B/B-UFSAR 3.7A-70 Shear Walls The results of the assessment are as follows:
Number of Walls Vertical Horizontal Acceptance Criteria Steel Steel ACI (0.9 Fy) 'cF= 3,500; Fy = 60,000 45 64 ACI (0.9 Fy) 'cF= 3,500; Fy = 67,000* 19 0 1.0 Fy 'cF= 3,500; Fy = 67,000* 1 0 1.035 Fy (M.F. = 0.966) 'cF= 3,500; Fy = 67,000* 0 1 65 65 *Actual average material strength for Rebar.
Concrete Beams and Slabs No unique design features were located.
Steel Columns The results of the assessment are as follows: Acceptance Criteria Number of Columns AISC Allowable 62 Fy = 50 AISC Allowable 4 Fy = 56* 66 *Actual average material strength for A577 structural steel.
B/B-UFSAR 3.7A-71Steel Beams The results of the assessment are as follows: Acceptance CriteriaNumberofBeams Lesser of 0.95 Fy or 1.6 AISC Allowable Fy = 36 189 Lesser of 0.95 Fy or 16 AISC Allowable Fy = 42.3* 52 1.0 Fy 1 Fy = 42.3* 242 *Actual average material strength for A36 structural steel.a.In reassessing the piping systems, all the pipingsubsystems or problems were identified and groupedaccording to system as shown in Table Q110.66-1.From the 319 piping problems listed, a representativesample of 40 piping problems were selected using thefollowing criteria:1.Selected all piping problems with Level Cdesign basis levels "Equation 9 NB/NC-3600"greater than 80% of the allowable limits.2.Selected piping problems ranging in size from3/4 inch to 48 inches nominal diameter.3.Selected piping problems from a range ofbuilding elevations to provide a diversity ofresponse spectra input.4.Selected at least one piping problem from eachsystem required for safe shutdown.Methodology In assessing the piping problems, the problems were analyzed for at least 30 modes or a range of frequency covering 33 Hz. The damping valves used were in accordance with Regulatory Guide 1.61 and the resultant stresses for service Level C were compared against the allowable stress limits of Articles NB/NC-3600 of ASME PBC Code Section III.
B/B-UFSAR 3.7A-72 Table Q110.66-2 contains a comparison of the highest stress points in a given piping problem, the function of the line, location, elevation, and allowable stress limits. b. A total of 89 valves were located in the 40 selected piping subsystems. These valves were assessed as follows: 1. 61 valves had to meet allowable acceleration limits in each mutually perpendicular direction (X,Y,Z). 2. 28 valves had to meet allowable acceleration limits in horizontal and vertical direction. Table Q110.66-3 contains a comparison of the inline valve accelerations using the design basis and Marble Hill response spectra for the SSE event. The valve types, sizes, and allowable accelerations are also shown in Table Q110.66-3. c. There are a total of 875 supports on the 40 piping problems selected. Of these 875, only 242 supports had load increases. Table Q110.66-4 contains the comparison of the support loads resulting from the dynamic analysis using the design basis and Marble Hill response spectra for the SSE event. Also included is the support number and type, and the maximum load carrying capacity of the support for service Level C. d. The difference between the results of the dynamic analysis using the design basis and the Marble Hill response spectra for the SSE event is minimal. Figures Q110.66-1 through Q110.66-4 provide a qualitative measure of the overall margin to failure for the piping problems, valves, and supports based on the Marble Hill spectra.
B/B-UFSAR 3.7A-73TABLE Q110.66-1 BREAKDOWN OF PIPING SUBSYSTEMS* NUMBEROFSYSTEMPIPINGPROBLEMSMain Steam System 5 Main Feedwater System 18 Auxiliary Feedwater System 15 Emergency Diesel Generator System 20 Component Cooling System 64 Essential Service Water System 52 Chemical and Volume Control System 70 Borated Water System 3 Residual Heat Removal System 20 Reactor Coolant System 30 Chilled Water System 22 TOTAL319* Piping systems required for safe shutdown of the plant assumingthe occurrence of an safe shutdown earthquake event.
B/B-UFSAR 3.7A-74TABLE Q110.66-2 STRESS LEVELS FOR ASME CLASS PIPING LEVEL C EQ. 9 STRESS(1) (KS1) SUB-CODE BLDG. PIPE DESIGN MARBLE LEVEL C STRESSSYSTEMLINECLASS ELEVATIONELEMENT BASIS HILL LIMIT (2) (KS1)
B/B-UFSAR 3.7A-75TABLE Q110.66-2 (Cont'd) LEVEL C EQ. 9 STRESS(1) (KS1) SUB-CODE BLDG. PIPE DESIGN MARBLE LEVEL C STRESSSYSTEMLINECLASSELEVATIONELEMENT BASIS HILL LIMIT (2) (KS1)
B/B-UFSAR 3.7A-76TABLE Q110.66-2 (Cont'd) LEVEL C EQ. 9 STRESS(1) (KS1) SUB-CODE BLDG. PIPE DESIGN MARBLE LEVEL C STRESSSYSTEMLINECLASSELEVATIONELEMENT BASIS HILL LIMIT (2) (KS1)
B/B-UFSAR 3.7A-77TABLE Q110.66-2 (Cont'd) LEVEL C EQ. 9 STRESS(1) (KS1) SUB-CODE BLDG. PIPE DESIGN MARBLE LEVEL C STRESSSYSTEMLINECLASSELEVATIONELEMENT BASIS HILL LIMIT (2) (KS1)
B/B-UFSAR 3.7A-78TABLE Q110.66-2 (Cont'd) LEVEL C EQ. 9 STRESS(1) (KS1) SUB-CODE BLDG. PIPE DESIGN MARBLE LEVEL C STRESSSYSTEMLINECLASSELEVATIONELEMENT BASIS HILL LIMIT (2) (KS1)
B/B-UFSAR 3.7A-79TABLE Q110.66-2 (Cont'd) LEVEL C EQ. 9 STRESS(1) (KS1) SUB-CODE BLDG. PIPE DESIGN MARBLE LEVEL C STRESSSYSTEMLINECLASSELEVATIONELEMENT BASIS HILL LIMIT (2) (KS1)
B/B-UFSAR 3.7A-80 TABLE Q110.66-3 VALVE ACCELERATIONS ON ASME PIPING SUB- VALVE DESIGN BASIS SSE MARBLE HILL SSE ALLOWABLE SSE SYSTEM SIZE/TYPE H1 H2 V H1 H2 V H1 H2 V 1AF-02 3 in. Control (A.O) 0.564 1.333 0.955 0.562 1.185 0.718 3.0 2.5 3.0 3 in. Control (A.O) 0.906 1.264 1.059 0.679 1.090 0.774 3.0 2.5 3.0 3 in. Control (A.O) 1.393 1.575 0.992 1.061 1.272 0.753 3.0 2.5 3.0 3 in. Control (A.O) 1.545 1.057 1.211 1.029 0.833 0.869 3.0 2.5 3.0 1AF-05 6 in. Check 1.718 1.836 1.001 1.661 1.857 1.042 3.0 2.5 3.0 6 in. Gate 1.406 1.857 0.756 1.436 1.921 0.744 3.0 2.5 3.0 6 in. Control 1.459 1.983 0.566 1.520 2.051 0.504 3.0 2.5 3.0 6 in. Gate 1.578 2.111 0.777 1.633 2.177 0.675 3.0 2.5 3.0 6 in. Check 1.354 1.073 1.853 1.307 1.194 1.955 3.0 2.5 3.0 4 in. Globe 1.198 0.987 1.263 1.195 1.093 1.271 3.0 2.5 3.0 4 in. Check 1.315 1.534 1.269 1.250 1.561 1.302 3.0 2.5 3.0 4 in. Globe 1.678 1.467 0.887 1.592 1.370 0.896 3.0 2.5 3.0 1AF-06 6 in. Check 0.559 0.573 1.183 0.506 0.573 1.212 3.0 2.5 3.0 6 in. Gate (M) 0.569 1.149 1.149 0.516 0.580 1.160 3.0 2.5 3.0 4 in. Control (A.O) 0.500 0.634 0.922 0.455 0.626 0.777 3.0 2.5 3.0 6 in. Gate (M) 1.216 1.005 1.561 1.098 0.924 1.269 3.0 2.5 3.0 4 in. Nozl Check 0.956 0.869 1.539 0.897 0.829 1.365 3.0 2.5 3.0 4 in. Globe (MO) 1.143 0.997 0.882 0.883 0.973 0.780 3.0 2.5 3.0 4 in. Nozl Check 1.102 0.452 0.914 0.931 0.436 0.671 3.0 2.5 3.0 4 in. Globe Iso (MO) 1.341 0.448 0.876 1.034 0.452 0.635 3.0 2.5 3.0 1AF-07 6 in. Check 1.109 0.953 0.869 0.960 0.846 0.670 3.0 2.5 3.0 6 in. Gate (M) 0.833 0.717 0.911 0.726 0.649 0.728 3.0 2.5 3.0 6 in. Control (AO) 0.598 0.481 0.851 0.530 0.430 0.646 3.0 2.5 3.0 6 in. Gate (M) 0.714 0.570 1.117 0.624 0.501 0.825 3.0 2.5 3.0
B/B-UFSAR 3.7A-81TABLE Q110.66-3 (Cont'd) SUB-VALVEDESIGN BASIS SSE MARBLE HILL SSE ALLOWABLE SSE SYSTEMSIZE/TYPEH1 H2 VH1 H2V H1 H2 V 1AF07 6 in. Check 1.126 0.880 2.444 0.924 0.710 1.534 3.0 2.5 3.0 4 in. Globe (MO) 1.702 1.103 1.338 1.035 0.787 0.865 3.0 2.5 3.0 4 in. Check 0.856 0.641 0.936 0.719 0.547 0.685 3.0 2.5 3.0 4 in. Globe (MO) 1.568 0.663 0.855 1.171 0.564 0.622 3.0 2.5 3.0 1CC-42 3 in. Gate (M) 0.760 0.375 1.034 0.535 0.339 0.608 2.122.122.0 3 in. Gate (M) 0.430 1.943 1.470 0.816 1.051 0.798 2.122.122.0 1DO-16 3 in. Relief 0.478 0.853 0.657 0.497 0.841 0.600 1DO-17 3 in. Gate 0.277 0.347 0.825 0.293 0.341 0.608 3.0 2.5 3.0 3 in. Gate 0.258 0.312 0.816 0.289 0.324 0.592 3.0 2.5 3.0 3 in. Gate 0.945 0.453 0.857 0.819 0.412 0.739 3.0 2.5 3.0 3 in. Gate 0.289 0.477 0.812 0.315 0.419 0.623 3.0 2.5 3.0 1FW-12 3 in. Check 1.499 1.220 1.377 1.303 1.071 1.151 3.0 2.5 3.0 3 in. Control 0.987 0.857 0.990 0.874 0.789 0.803 3.0 2.5 3.0 3 in. Gate 1.482 1.229 0.696 1.232 1.066 0.517 3.0 2.5 3.0 3 in. Gate 0.886 0.849 0.561 0.763 0.744 1.465 3.0 2.5 3.0 3 in. Control 0.476 0.552 0.949 0.423 0.504 0.705 3.0 2.5 3.0 3 in. Gate 0.878 0.876 0.976 0.729 0.786 0.814 3.0 2.5 3.0 1SX-02 42 in. Butterfly 0.244 0.436 0.842 0.269 0.412 0.609 3.0 2.5 3.0 42 in. Butterfly 0.230 0.389 0.833 0.238 0.390 0.602 3.0 2.5 3.0 42 in. Butterfly 0.412 0.799 0.983 0.363 0.696 0.758 3.0 2.5 3.0 42 in. Butterfly 0.274 0.644 1.012 0.290 0.554 0.738 3.0 2.5 3.0 B/B-UFSAR 3.7A-82 TABLE Q110.66-3 (Cont'd) SUB- VALVE DESIGN BASIS SSE MARBLE HILL SSE ALLOWABLE SSE SYSTEM SIZE/TYPE H1 H2 V H1 H2 V H1 H2 V 1SX-02 42 in. Butterfly 0.525 0.591 1.157 0.454 0.459 0.077 3.0 2.5 3.0 42 in. Butterfly 0.570 0.602 1.295 0.559 0.470 0.978 3.0 2.5 3.0 24 in. Butterfly 0.323 0.309 0.991 0.321 0.321 0.839 3.0 2.5 3.0 24 in. Butterfly 0.798 0.604 1.164 0.699 0.529 1.004 3.0 2.5 3.0 0.75 in. Globe 0.672 0.284 1.171 0.630 0.305 0.845 3.0 2.5 3.0 24 in. Butterfly 0.599 0.396 0.966 0.593 0.370 0.714 3.0 2.5 3.0 30 in. Butterfly 0.304 1.253 0.896 0.328 0.937 0.650 3.0 2.5 3.0 1SX-05 16 in. Butterfly 0.518 0.449 1.036 0.468 0.639 0.875 3.0 2.5 3.0 1SX-13 10 in. Butterfly 0.768 0.450 1.685 0.737 0.569 2.055 3.0 2.5 3.0 10 in. Butterfly 0.802 0.451 1.534 0.788 0.568 1.899 3.0 2.5 3.0 10 in. Butterfly 0.887 0.453 1.480 0.890 0.565 1.732 3.0 2.5 3.0 10 in. Butterfly 1.162 0.753 1.189 1.255 0.796 1.149 3.0 2.5 3.0 1WO-32 3 in. Globe 0.746 0.641 1.191 0.725 0.861 1.043 3.0 2.5 3.0 3 in. Globe 0.850 0.925 1.162 0.906 1.149 1.032 3.0 2.5 3.0 3 in. Globe 0.862 1.137 1.033 0.921 1.624 0.881 3.0 2.5 3.0 3 in. Globe 0.829 1.16691.324 0.793 2.252 1.254 3.0 2.5 3.0 B/B-UFSAR 3.7A-83 REVISION 1 - DECEMBER 1989 TABLE Q110.66-3 (Cont'd) DESIGN BASIS SSE MARBLE HILL SSE ACCELERATIONS ALLOWABLE ACCELERATIONS VALVE SIZE (g) (g) (g) SUBSYSTEM LINE AND TYPE H V H V H V 1RH02 RHR System 12 in. Gate 2.7 2.0 2.1 1.5 3.0 2.0 12 in. Gate 4.2 0.7 3.3 0.9 3.0 2.0(1) 12 in. Gate 4.6 2.6 3.8 1.9 3.0 2.0(1) 12 in. Gate 2.0 1.9 2.4 1.4 3.0 2.0 1SI04 Accumulator 10 in. Gate 3.2 1.8 3.5 1.9 4.2 3.0 System 1SI09 Accumulator 10 in. Gate 2.3 0.9 2.1 0.7 4.2 3.0 System 1CV06 CVCS 3 in. Gate 2.9 1.4 2.6 1.6 3.0 2.0 3 in. Globe 3.6 1.9 2.8 2.3 6.0 4.0 1CV60 Excess 1 in. Globe 0.1 0.1 0.9 0.5 3.0 2.0 Letdown 1 in. Globe 0.1 0.1 0.5 0.6 3.0 2.0 1CC25 Component 2 in. Globe 2.3 2.1 2.8 2.4 8.5 4.0 Cooling 2 in. Globe 2.4 1.4 3.7 1.7 8.5 4.0 4 in. Gate 2.0 1.5 2.5 1.4 3.0 2.0 1RC21 Sample 3/4 in. Globe 0.1 0.0 0.7 0.5 3.0 2.0 Line 1CV36 RCP Seal 3/4 in. Globe 2.36 3.52 2.2 3.27 8.5 4 (1) Higher limits to be qualified.
B/B-UFSAR 3.7A-84 REVISION 1 - DECEMBER 1989 TABLE Q110.66-3 (Cont'd) DESIGN BASIS SSE MARBLE HILL SSE ACCELERATIONS ALLOWABLE ACCELERATIONSVALVE SIZE(g)(g)(g)SUBSYSTEMLINEAND TYPEHVHVHV1CC21 Component 3 in. Globe 1.6 1.3 0.8 0.6 6 4 1CC22 Cooling 3 in. Globe 4.1 1.9 3.0 0.9 6 4 1CC23 3 in. Globe 4.2 2.9 1.7 1.7 6 4 3 in. Globe 4.7 3.6 2.2 2.3 6 4 3 in. Control 3.3 0.6 1.9 0.3 3(1) 2 1SX06 Service 10 in. Butterfly 2.4 0.38 2.42 1.31 3 2 Water 10 in. Butterfly 0.62 0.38 1.65 1.16 3 2 10 in. Butterfly 0.16 0.34 0.35 1.24 3 2 10 in. Butterfly 0.66 0.26 1.55 1.0 3 2 1SX08 Service 10 in. Butterfly 1.12 0.3 1.66 1.16 3 2 Water 10 in. Butterfly 1.68 0.54 2.11 1.29 3 2 10 in. Butterfly 1.14 0.42 2.26 1.43 3 2 10 in. Butterfly 1.46 0.68 1.61 1.38 3 2 (1) Higher limits to be qualified.
B/B-UFSAR 3.7A-85TABLE Q110.66-4 PIPING SUPPORTS DESIGN BASIS MARBLE HILL PERCENT SUPPORT SUBSYSTEMSUPPORTLOADLOADINCREASE LOAD CAPACITY1AF-021AF02006X6496774.380001AF02016X47678665.580001AF02009X66574311.280001P Guide184018842.4-1AF02013X53960612.480001AF-051AF05061R474847800.6760001AF05002X7297624.596001AF05058R59565910.722501AF05004R152515702.924101AF05006R5886073.215001AF05007R4724965.115001AF05008R4064070.24 15001AF05010R2963011.715001AF05012R4304545.615001AF05014R3613825.815001AF05015R3683772.415001AF05017R3633784.115001AF05018R3633722.415001AF05020R3633763.615001AF05021R3633897.215001AF05023R3643742.715001AF05025R3633660.81500___________________
Notes: R - Rigid in Y-Direction.S -SnubberX - Rigid in Horizontal Direction. G - Guides B/B-UFSAR 3.7A-86 TABLE Q110.66-4 (Cont'd) DESIGN BASIS MARBLE HILL PERCENT SUPPORT SUBSYSTEM SUPPORT LOAD LOAD INCREASE LOAD CAPACITY 1AF-05 1AF05026R 376 393 4.5 1500 1AF05028R 397 422 6.3 1500 1AF05029R 387 408 6.4 1500 1AF11002X 3853 2871 0.6 9600 1AF11003X 3023 3068 1.5 9600 1AF11021R 580 650 12.1 1500 1AF11005X 764 817 6.9 9600 1AF11006X 864 900 4.3 9600 1AF05072R 1625 1648 1.4 2410 1AF05040R 404 442 9.4 1500 1AF05045R 362 421 16.3 1500 1AF05048R 361 390 8.0 1500 1AF05049R 369 392 6.2 1500 1AF05051R 377 407 7.9 1500 1AF05053R 376 403 7.2 1500 1AF05065R 380 470 23.7 870 1AF12003X 553 568 2.7 9600 1AF12002X 544 557 2.4 9600 1AF12001X 506 2109 316.7 9600 1AF12005X 1360 1458 7.2 9600 1AF12004R 532 543 2.1 1500 1AF12006X 590 720 22.0 9600 1AF05060S NOT AVAILABLE 1AF-05 1AF05063S 3936 4035 2.5 8610 1AF05064S 2428 2510 3.4 8610 1AF05009S 4134 4231 2.3 8610 1AF-06 1AF06001R 1407/-61 1415/-69 0.6 2410 1AF06013R 920/-142 937/-159 1.8 NON-STANDARD 1AF06017R 820 828 0.9 930 1AF06025X 736/-661 972 32 9600 1AF06026X 494/-580 618 4.5 9600 B/B-UFSAR 3.7A-87 TABLE Q110.66-4 (Cont'd) DESIGN BASIS MARBLE HILL PERCENT SUPPORT SUBSYSTEM SUPPORT LOAD LOAD INCREASE LOAD CAPACITY 1AF06032R 524 525 0.2 930 1AF06014R 589 627 6.5 1500 1AF06028X 530/-352 566 6.8 9600 1AF-07 1AF07026X 372 394 5.9 9600 1AF07009X 630 639 1.4 9600 1AF07016R 729/-161 749 2.7 1500 1AF07024X 335 378 12.8 9600 1DO-16 1DO16004X 68/-1163 76/-1171 0.7 NON-STANDARD 1DO16006G 41/-20 51/-31 26.8 240 1DO16006G 50/-119 51/-121 1.6 3010 1DO16008G 44/-50 47/-53 6.0 240 155/-93 155/ 3010 1DO-17 1DO17002X 650 714 9.8 3010 1DO17008G 212 224 5.6 6020 175 165 480 1SX-02 1SX02031R 32740 33045 0.93 33500 1SX02015S 33906 34172 0.080 50000 1SX-05 1SX05009X 6481 6528 0.72 9600 1SX00510X 4953 5039 1.73 9600 1SX05011X 6007 6113 1.76 9600 1SX-13 1SX13001X 3334 3758 12.7 4500 1SX13002R 3261 3632 11.4 2710/ROD 1SX13014R 4283 4813 12.4 3770/ROD 1SX13016X 2216 2250 1.5 4500 1SX13021R 1794 1905 6.2 1810/ROD B/B-UFSAR 3.7A-88 TABLE Q110.66-4 (Cont'd) DESIGN BASIS MARBLE HILL PERCENT SUPPORT SUBSYSTEM SUPPORT LOAD LOAD INCREASE LOAD CAPACITY 1SX-13 1SX13011X 1592 1629 2.3 3240 1SX13020X 3955 4140 1.7 2710/ROD 1SX13019R 1806 1869 3.4 1130/ROD 1SX13018X 3139 3459 10.2 4500 1SX-34 1SX34001X 1731 1769 2.2 2250 1SX34004X 840 951 13.2 11630 1SX34007R 1437 1591 10.7 2250 1SX34012X 542 721 33.0 1500 1SX34008R 1922 2374 23.5 2710 1SX34009X 594 831 39.9 4500 1SX34013X 927 1210 30.5 4500 1SX34010R 3773 5144 36.3 11630 1SX34011X 1268 1714 35.2 11630 1SX-63 1SX63001X 1110 1199 8.0 1500 1SX63002R 3102 3231 4.1 4500 1SX63003X 2018 2129 5.5 4500 1SX63004X 1507 1579 4.8 2250 1SX63005R 1733 1787 3.1 2710 1SX6300SX 942 945 0.21 1500 1SX63007R 1671 1726 3.3 4500 1SX63008X 627 655 4.5 1500 1WO-32 1WO32002X 1495 1531 2.4 NON-STANDARD 1WO32993R 938 956 0.95 1502 1WO32004X 126 134 6.3 3710 1WO32005R 986 994 0.81 1502 1WO32006X 304 373 22.7 3710 1WO32008X 1WO32007X 980 1001 2.1 1500 1WO32010X 326 431 32.2 NON-STANDARD 1WO32012X 370 456 23.2 3710 B/B-UFSAR 3.7A-89 TABLE Q110.66-4 (Cont'd) DESIGN BASIS MARBLE HILL PERCENT SUPPORT SUBSYSTEM SUPPORT LOAD LOAD INCREASE LOAD CAPACITY 1WS32013R 662 670 1.2 1502 1WO32014X 986 1034 4.8 3710 1WO32015R 572 591 3.3 870 1WO32016R 327 355 8.6 3710 1WO32017R 874 925 5.8 1500 1WO32018X 265 277 4.5 870 1WO32020X 342 361 5.5 870 1WO32023X 538 575 6.9 870 1WO32037X 1164 1170 0.5 6000 1WO32032R 701 720 2.7 1502 1W032038X 397 401 2.5 3710 1WS32039R 994 1043 4.9 2407 1WO32040R 573 599 4.5 NON-STANDARD 1WO32042X 569 601 5.6 870 1WO32044R 1189 1221 2.7 2407 1WO32045X 223 281 26.0 6000
B/B-UFSAR 3.7A-90 REVISION 1 - DECEMBER 1989 TABLE Q110.66-4 (Cont'd) SUPPORT TYPE DESIGN MARBLE HILL SYSTEM NUMBER (S,R,F) BASIS LOAD LOAD CAPACITY 1SI04 SI4-004S S 3.1 3.8 8.61 ADD-SNUB S 8.9 9.5 NSC (New load accepted) SI4-009S S 9.2 9.7 13.96 ADD-Y-SN S 10.2 10.7 NSC (New load accepted) ADD-H-SN S 4.6 5.5 NSC (New load accepted) SI4-017S S 4.7 4.9 8.61 1SI09 SI9-020S S 9.2 10.7 70.35 SI9-021S S 6.1 7.3 8.61 ADD-2-SK R 7.2 7.7 NSC (New load accepted) SI9-024S S 12.9 15.8 20.1 ADD-X-DS S 8.0 9.2 NSC (New load accepted) SI9-015S S 2.9 3.1 8.61 SI9-14R F 5.1 5.5 NSC (New Load accepted) SI9-04S S 6.0 7.4 8.61 1CV06 1CV06-15 R 1.1 1.2 24.84 1CV06-14 S 1.2 1.4 20.1 1CV06-08 S 2.0 2.3 8.61 1CV06-10 S 1.0 1.2 2.067 1CV02 1CV02003 S 2.1 2.5 8.6 1CV02010 S 1.2 1.4 2.1 1CV02005 S 0.5 0.6 0.6 1CV02111 F 2.6 2.9 NSC (New load accepted) 1RH02 RHR062R F 15.5 26.7 NSC (New load accepted) RHR058S S 18.4 18.5 20.1 ADDY R 9.0 9.5 NSC (New load accepted)
B/B-UFSAR 3.7A-91 REVISION 1 - DECEMBER 1989 TABLE Q110.66-4 (Cont'd) SUPPORT TYPE DESIGN MARBLE HILL SYSTEM NUMBER (S,R,F) BASIS LOAD LOAD CAPACITY 1CV09 005 S 0.3 0.4 0.9 045 S 0.4 0.5 0.5 047 S 0.2 0.3 0.9 013 R 0.5 0.6 2.3 014 R 0.2 0.3 1.2 016 R 0.2 0.3 1.5 019 R 0.3 0.4 2.3 020 R 0.4 0.5 1.3 030 S 0.4 0.5 2.1 024 F 0.4 0.6 NSC (New load accepted) 035 S 0.2 0.3 0.5 036 R 0.4 0.5 0.9 039 R 0.1 0.2 0.9 REIRC02ACA F 0.4 0.5 NSC (New load accepted) 1CV60 VALVE 1 R 0.2 0.3 NSC (New load accepted) 1SD04 004S S 0.18 0.23 0.5 RE F 0.25 0.26 NSC (New load accepted) 1SI02 ANCHOR 1 F 7.0 7.1 NSC (New load accepted) 1CC25 006R F 1.4 1.6 NSC (New load accepted) 010R F 0.7 1.0 NSC (New load accepted) 013R F 1.0 1.1 NSC (New load accepted) 016X R 2.5 2.8 6.0 017R R 1.0 1.2 1.5 009X R 0.7 0.9 0.87 1RB-68 F 2.4 2.7 NSC (New load accepted) 1MS06 0045B S 47.2 48.1 On Hold 001X 41.6 44.3 On Hold B/B-UFSAR 3.7A-92 REVISION 3 - DECEMBER 1991 TABLE Q110.66-4 (Cont'd) SUPPORT TYPE DESIGN MARBLE HILL SYSTEM NUMBER (S,R,F) BASIS LOAD LOAD CAPACITY 1SX08 SX08001R R 6.9 9.0 9.6 002R R 6.0 7.2 9.6 003R R 4.7 6.3 6.0 004R R 3.0 4.6 6.0 005R R 4.6 5.7 6.0 019X R 5.2 6.4 14.1 006R R 3.4 5.0 NSC (New load accepted) 015X R 2.9 3.6 14.1 007R1 R 4.3 5.6 NSC (New load accepted) 008R1 R 4.8 7.0 NSC (New load accepted) 009R1 R 4.0 6.7 NSC (New load accepted) 010R1 R 8.9 10.5 NSC (New load accepted) 011R1 R 5.1 10.5 NSC (New load accepted) 013R1 R 4.6 12.0 NSC (New load accepted) 014R1 R 3.6 4.3 NSC (New load accepted) 021S S 2.3 2.8 8.6 020R R 9.7 11.5 19.2 017S S 1.1 1.4 2.1 016X R 2.1 2.4 14.1 025R R 8.0 9.2 19.2 1SD03 1SD03008S S 0.2 0.3 0.5 019R R 0.3 0.4 0.9 013X S 0.3 0.4 0.5 014X R 0.1 0.2 0.9 1CC21, 22, 23 1CC22005 R 0.9 1.1 NSC (New load accepted) 023 R 0.3 0.5 1.5 034 F 0.4 0.5 NSC (New load accepted) 1SX06 SX0601R R 3.2 4.3 4.5 02R R 2.4 3.8 NSC (New load accepted) 03R R 5.1 6.6 NSC (New load accepted)
B/B-UFSAR 3.7A-93 REVISION 1 - DECEMBER 1989 TABLE Q110.66-4 (Cont'd) SUPPORT TYPE DESIGN MARBLE HILL SYSTEM NUMBER (S,R,F) BASIS LOAD LOAD CAPACITY 04R R 6.7 13.6 NSC (New load accepted) 05R R 5.3 7.4 NSC (New load accepted) 06R R 4.8 7.6 NSC (New load accepted) 07R R 4.8 7.9 NSC (New load accepted) 08R R 4.3 7.0 NSC (New load accepted) 09R R 4.8 6.8 NSC (New load accepted) 20R R 4.7 6.8 NSC (New load accepted) 11R R 2.9 3.2 3.0 (Faulted 4.0) 14X R 3.0 4.4 NSC (New load accepted) 15R R 1.0 1.6 1.7 34R R 3.6 6.1 5.0 (Faulted 7.1) 36S S 3.0 4.3 8.6 35X R 3.6 4.5 6.0 37X R 1.9 2.6 3.7 22R R 5.7 6.4 6.6 24X R 1.2 1.8 7.8 23X R 3.3 4.7 7.8 25X R 1.3 1.7 3.1 27X R 0.4 0.6 3.7 26X R 2.5 2.8 3.7 16R R 5.3 6.4 6.6 18R R 3.4 7.8 8.0 (Faulted 8.0) 17X R 3.2 5.6 14.0 19X R 1.2 2.5 2.3 (Faulted 3.0) 20X R 0.6 1.0 2.2 21X R 2.6 3.5 6.0 28R R 4.5 6.5 6.2 (Faulted 8.8) 29X R 4.2 5.2 7.8 30X R 4.5 5.7 7.8 31X R 3.4 3.9 7.8 33X R 4.2 4.6 7.8 32X R 2.6 3.0 7.8
B/B-UFSAR 3.7A-94 REVISION 1 - DECEMBER 1989 TABLE Q110.66-4 (Cont'd) SUPPORT TYPE DESIGN MARBLE HILL SYSTEM NUMBER (S,R,F) BASIS LOAD LOAD CAPACITY 1RY05 1RY05003 S 32.6 38.5 70.4 008 R 21.8 22.0 37.7 010 S 17.3 18.3 20.1 1CV03 1CV03002 S 1.3 1.5 2.1 015 S 0.8 0.9 0.9 011 R 0.9 1.1 4.5
B/B-UFSAR 3.7A-95 3.7A.4 Byron/Braidwood Detailed Review of the Seismic Design of One Piping Subsystems NRC Question 110.67 requested a detailed comparison of the design basis spectra versus the new (R.G. 1.60) seismic spectra for one selected piping subsystem. The response evaluated the 15X-13 piping problem for the essential service water system. The comparison between the Marble Hill (Regulatory Guide 1.60) spectra and the Byron/Braidwood design basis spectra is provided. QUESTION 110.67 "Provide a detailed discussion of one of the selected piping subsystems and include a comparison of the design basis spectra used versus the new seismic spectra applicable to the subsystem showing those frequencies where the new seismic spectra are not bounded by the design basis spectra. Include a discussion on the modal frequencies and participation factors of the selected piping subsystem. Show what the resulting effects of those frequencies where the new seismic spectra are not bounded by the design basis spectra are on the piping stresses, support loads, and valve accelerations."
RESPONSE Piping problem 1SX-13, Figure Q110.67-1, was selected for the detailed discussion for the following reasons: a. It spans through three different auxiliary building elevations. b. It contains four valves at different locations.
- c. The Marble Hill response spectra is not bounded by the design basis spectra at the higher frequencies. See Figures Q110.67-2, Q110.67-3, and Q110.67-4. d. It is a 10-inch pipe, which is a good representative size for a study of this nature. Table Q110.67-1 contains the dynamic characteristics of the piping system, specifically the modal periods and participation factors of piping subsystem 1SX-13. Table Q110.67-2 contains the highest participation factors for four modes in each direction, the corresponding modal periods and acceleration values for both the design basis and the Marble Hill response spectra.
Table Q110.67-3 is a listing of the seismic stresses using the design basis response spectra.
B/B-UFSAR 3.7A-96 Table Q110.67-4 is a listing of the seismic stresses using the Marble Hill response spectra. Table Q110.67-5 provides a comparison of the valve accelerations resulting from both analyses versus the allowable acceleration values.
Table Q110.67-6 provides a comparison of the support loads resulting from both analyses versus the allowable load carrying capacity of the supports for service Level C limits. The results of these comparisons indicate there is enough margin left in the pipe stresses, valve accelerations, and support loads for service C limits using the Marble Hill response spectra.
B/B-UFSAR 3.7A-97 TABLE Q110.67-1 DYNAMIC CHARACTERISTICS OF THE PIPING SYSTEM 4391-00 B/B-1 1SX-1 MODAL PERIODS PARTICIPATION FACTORS MODE (SEC) X Y Z 1 0.18823 1.047 0.194 -2.111 2 0.14355 0.794 0.104 -1.497 3 0.10304 2.024 -0.179 -1.027 4 0.07878 1.317 0.678 -0.315 5 0.06669 -0.293 1.865 -0.231 6 0.05508 -1.950 -0.880 -0.816 7 0.04535 0.566 0.490 0.890 8 0.03987 -0.003 1.233 0.680 9 0.03529 -0.810 0.165 -0.356 10 0.03362 -1.534 -0.828 0.393 11 0.03265 0.194 -0.641 0.961 12 0.03072 -0.429 0.544 0.363 13 0.02786 0.013 1.281 -0.061 14 0.02517 1.685 -1.432 0.395 15 0.02333 -0.248 -0.475 0.251 16 0.02169 0.121 0.804 -1.048 17 0.02020 -0.942 -0.032 0.078 18 0.01887 -0.274 -0.351 -1.353 19 0.01702 0.144 0.994 0.116 20 0.01633 0.283 0.344 0.235 21 0.01622 0.335 -1.019 0.021 22 0.01557 -0.393 -0.520 0.107 23 0.01447 0.027 0.497 0.626 24 0.01387 1.103 0.744 0.177 25 0.01339 0.469 -0.464 0.665 26 0.01314 1.132 -0.162 0.685 27 0.01295 -0.192 0.033 0.405 28 0.01248 0.559 -1.797 0.207 29 0.01211 0.013 -1.282 0.386 30 0.01196 -0.255 0.402 -0.055 B/B-UFSAR 3.7A-98 REVISION 9 - DECEMBER 2002 TABLE Q110.67-2 PARTICIPATION FACTORS, MODAL PERIODS, AND ACCELERATION VALUES FOR DESIGN BASIS AND MARBLE HILL SPECTRA
MODAL PARTICIPATION DESIGN BASIS MARBLE HILL MODE PERIOD FACTOR ACCELERATION (g) ACCELERATION (g) X-DIRECTION 3 0.10304 2.024 1.50 1.49 6 0.05508 -1.950 0.85 0.90 10 0.03362 -1.534 0.48 0.52 14 0.02517 1.685 0.48 0.52 Y-DIRECTION 5 0.06669 1.865 3.00 4.20 8 0.03987 1.233 1.00 1.10 13 0.02786 1.281 0.95 0.90 14 0.02517 -1.432 0.95 0.85 Z-DIRECTION 1 0.18823 -2.111 1.90 2.15 2 0.14355 -1.497 1.60 1.60 16 0.02169 -1.048 0.42 0.61 18 0.01887 -1.353 0.42 0.61 B/B-UFSAR 3.7A-99 TABLE Q110.67-3 SEISMIC STRESSES USING DESIGN BASIS SPECTRA FACTOR DISPLACEMENTS IN INCHES NODE TYPE I STRESS IN PSI X Y Z AXIS 5 3 1.00 2418. .000 .000 .000 GLOB 10A 9 2.61 6297. .000 .000 .000 GLOB 10B 9 2.61 4753. .026 .018 .004 GLOB 13 1 1.00 872. .122 .030 .009 GLOB 15A 9 2.61 5105. .241 .044 .023 GLOB 15B 9 2.61 5734. .256 .045 .023 GLOB 17 1 1.00 2591. .235 .030 .023 GLOB 18 1 1.00 2539. .215 .020 .000 SKEW 20 1 1.00 2445. .168 .000 .023 GLOB 23 1 1.00 2333. .087 .038 .023 GLOB 25 1 1.00 4354. .000 .068 .023 GLOB 33 7 1.90 6819. .078 .071 .023 GLOB 35 7 1.90 5745. .098 .069 .023 GLOB 40 7 1.90 4830. .121 .065 .023 GLOB 45 7 1.90 4439. .142 .061 .023 GLOB 53 7 1.90 7416. .195 .049 .023 GLOB 55A 9 2.61 11459. .211 .045 .023 GLOB 55B 9 2.61 12100. .207 .022 .024 GLOB 57 1 1.00 4106. .207 .000 .100 GLOB 60 1 1.00 2956. .207 .031 .290 GLOB 65 1 1.00 2690. .208 .049 .658 GLOB 70A 9 2.61 6764. .208 .043 .946 GLOB 70B 9 2.61 6599. .000 .037 .929 SKEW 78 1 1.00 1517. .114 .037 .474 GLOB 80A 9 2.61 11560. .023 .037 .061 GLOB 80B 9 2.61 14156. .000 .001 .000 GLOB 85 4 2.10 11543. .000 .000 .000 GLOB 90A 9 2.61 3539. .171 .053 .018 GLOB 90B 9 2.61 2983. .142 .047 .020 GLOB 95A 9 2.61 3742. .043 .006 .020 GLOB B/B-UFSAR 3.7A-100 TABLE Q110.67-3 (Cont'd) FACTOR DISPLACEMENTS IN INCHES NODE TYPE I STRESS IN PSI X Y Z AXIS 95B 9 2.61 4068. .025 .000 .021 GLOB 100A 9 2.61 4108. .018 .002 .021 GLOB 100B 9 2.61 4625. .000 .005 .022 GLOB 105 1 1.00 1432. .069 .008 .022 GLOB 108 1 1.00 1880. .119 .000 .000 SKEW 110 1 1.00 1951. .138 .020 .022 GLOB 115A 9 2.61 4008. .129 .037 .022 GLOB 115B 9 2.61 3313. .113 .035 .020 GLOB 120A 9 2.61 3833. .019 .015 .001 GLOB 120B 9 2.61 4768. .000 .000 .000 GLOB 125 3 1.00 1793. .000 .000 .000 GLOB 130A 9 2.61 6507. .072 .073 .026 GLOB 130B 9 2.61 5563. .098 .066 .031 GLOB 135 1 1.00 1868. .107 .051 .031 GLOB 140 7 1.90 4328. .118 .026 .031 GLOB 145A 9 2.61 6618. .120 .018 .031 GLOB 145B 9 2.61 6753. .109 .000 .037 GLOB 150 1 1.00 1759. .109 .021 .149 GLOB 155 1 1.00 1763. .109 .032 .326 GLOB 160A 9 2.61 4640. .109 .027 .461 GLOB 160B 9 2.61 3807. .100 .022 .444 GLOB 165 1 1.00 2796. .000 .022 .300 SKEW 170A 9 2.61 6574. .015 .022 .036 GLOB 170B 9 2.61 8632. .000 .000 .000 GLOB 175 4 2.10 7279. .000 .000 .000 GLOB 178 1 1.00 0. .131 .043 .021 GLOB
B/B-UFSAR 3.7A-101 TABLE Q110.67-3 (Cont'd) FACTOR STRESS IN PSI DISPLACEMENTS IN INCHES NODE TYPE I BRANCH RUN X Y Z 30 WDT 1.97 5520. 8040. .063 .073 .023 30 UFT 5.28 14821. 21589. .063 .073 .023 30 RFT 5.28 14821. 21589. .063 .073 .023 50 WDT 1.97 4907. 7135. .180 .053 .023 50 UFT 5.28 13175. 19158. .180 .053 .023 50 RFT 5.28 13175. 19158. .180 .053 .023 93 WDT 1.97 1. 2145. .131 .043 .020 93 UFT 5.28 4. 5761. .131 .043 .020 93 RFT 5.28 4. 5761. .131 .043 .020
B/B-UFSAR 3.7A-102 TABLE Q110.67-4 SEISMIC STRESSES USING MARBLE HILL SPECTRA FACTOR DISPLACEMENTS IN INCHES NODE TYPE I STRESS IN PSI X Y Z AXIS 5 3 1.00 2367. .000 .000 .000 GLOB 10A 9 2.61 6166. .000 .000 .000 GLOB 10B 9 2.61 4697. .025 .018 .004 GLOB 13 1 1.00 860. .119 .030 .009 GLOB 15A 9 2.61 5090. .236 .043 .022 GLOB 15B 9 2.61 5695. .252 .044 .023 GLOB 17 1 1.00 2637. .233 .030 .023 GLOB 18 1 1.00 2683. .213 .020 .000 SKEW 20 1 1.00 2793. .168 .000 .023 GLOB 23 1 1.00 2415. .088 .044 .023 GLOB 25 1 1.00 4482. .000 .081 .023 GLOB 33 7 1.90 7181. .083 .087 .023 GLOB 35 7 1.90 6046. .104 .085 .023 GLOB 40 7 1.90 5048. .129 .081 .023 GLOB 45 7 1.90 4599. .151 .077 .023 GLOB 53 7 1.90 8058. .206 .064 .023 GLOB 55A 9 2.61 12231. .224 .059 .023 GLOB 55B 9 2.61 12819. .219 .028 .025 GLOB 57 1 1.00 4522. .219 .000 .106 GLOB 60 1 1.00 3390. .220 .040 .307 GLOB 65 1 1.00 2984. .220 .058 .698 GLOB 70A 9 2.61 7170. .220 .046 1.003 GLOB 70B 9 2.61 7048. .000 .040 .985 SKEW 78 1 1.00 1616. .120 .040 .502 GLOB 80A 9 2.61 12255. .025 .040 .065 GLOB 80B 9 2.61 15010. .000 .001 .000 GLOB 85 4 2.10 12238. .000 .000 .000 GLOB 90A 9 2.61 4211. .182 .068 .018 GLOB 90B 9 2.61 3047. .151 .059 .020 GLOB 95A 9 2.61 4059. .045 .007 .020 GLOB B/B-UFSAR 3.7A-103 TABLE Q110.67-4 (Cont'd) FACTOR DISPLACEMENTS IN INCHES NODE TYPE I STRESS IN PSI X Y Z AXIS 95B 9 2.61 4385. .026 .000 .020 GLOB 100A 9 2.61 4354. .018 .002 .021 GLOB 100B 9 2.61 4825. .000 .005 .021 GLOB 105 1 1.00 1459. .068 .008 .022 GLOB 108 1 1.00 1865. .117 .000 .000 SKEW 110 1 1.00 1917. .136 .020 .021 GLOB 115A 9 2.61 3969. .126 .036 .021 GLOB 115B 9 2.61 3299. .111 .034 .020 GLOB 120A 9 2.61 3802 .018 .015 .001 GLOB 120B 9 2.61 4706. .000 .000 .000 GLOB 125 3 1.00 1769. .000 .000 .000 GLOB 130A 9 2.61 6830. .078 .088 .026 GLOB 130B 9 2.61 5926. .105 .080 .031 GLOB 135 1 1.00 1990. .113 .063 .031 GLOB 140 7 1.90 4399. .122 .032 .031 GLOB 145A 9 2.61 6674. .123 .022 .031 GLOB 145B 9 2.61 6833. .112 .000 .036 GLOB 150 1 1.00 1820. .112 .022 .150 GLOB 155 1 1.00 1835. .112 .033 .329 GLOB 160A 9 2.61 4783. .112 .028 .465 GLOB 160B 9 2.61 3872. .102 .022 .447 GLOB 165 1 1.00 3074. .000 .022 .302 SKEW 170A 9 2.61 6637. .016 .022 .036 GLOB 170B 9 2.61 8713. .000 .000 .000 GLOB 175 4 2.10 7252. .000 .000 .000 GLOB 178 1 1.00 0. .140 .054 .021 GLOB
B/B-UFSAR 3.7A-104 TABLE Q110.67-4 (Cont'd) FACTOR STRESS IN PSI DISPLACEMENTS IN INCHES NODE TYPE I BRANCH RUN X Y Z 30 WDT 1.97 5730. 8447. .066 .088 .023 30 UFT 5.28 15386. 22682. .066 .088 .023 30 RFT 5.28 15386. 22682. .066 .088 .023 50 WDT 1.97 5809. 7917. .191 .068 .023 50 UFT 5.28 15598. 21258. .191 .068 .023 50 RFT 5.28 15598. 21258. .191 .068 .023 93 WDT 1.97 1. 2180. .139 .054 .020 93 UFT 5.28 4. 5854. .139 .054 .020 93 RFT 5.28 4. 5854. .139 .054 .020
B/B-UFSAR 3.7A-105 TABLE Q110.67-5 VALVE ACCELERATION ON ASME PIPING SUB- VALVE DESIGN BASIS SSE MARBLE HILL SSE ALLOWABLE SSE SYSTEM SIZE/TYPE H1 H2 V H1 H2 V H1 H2 V 1SX-13 10 in. Butterfly 0.768 .450 1.685 0.737 .569 2.055 3.0 2.5 3.0 10 in. Butterfly 0.802 .451 1.534 0.788 .568 1.899 3.0 2.5 3.0 10 in. Butterfly 0.887 .453 1.480 0.890 .565 1.732 3.0 2.5 3.0 10 in. Butterfly 1.162 .753 1.189 1.255 .796 1.149 3.0 2.5 3.0
B/B-UFSAR 3.7A-106 TABLE Q110.67-6 PIPING SUPPORTS SUBSYSTEM SUPPORT DESIGN BASIS LOAD MARBLE HILL LOAD PERCENT INCREASE SUPPORT LOAD CAPACITY 1SX-13 1SX13011X 1592 1692 2.3 3240 1SX13020X 3955 4140 1.7 2710/ROD 1SX13019R 1806 1869 3.4 1130/ROD 1SX13018X 3139 3459 10.2 4500 1SX13001X 3334 3758 12.7 4500 1SX13002R 3261 3632 11.4 2710/ROD 1SX13014R 4283 4813 12.4 3770/ROD 1SX13016X 2216 2250 1.5 4500 1SX13021R 1794 1905 6.2 1810/ROD
B/B-UFSAR 3.7A-107 3.7A.5 Byron Site Specific Seismic Concerns In NRC Question 130.9, it was requested that the soil-structure interaction evaluation for the Byron river screen house include the half-space lumped spring and mass method as well as the finite element method used for the Byron design. The response provided the soil-structure interaction responses using the half-space lumped parameter approach.
NRC Question 130.9a requested a quantitative assessment of the impact of enveloping the most severe responses of the two methods used in the response to Question 130.9 on the design of the Byron river screen house. The response to Question 130.9a outlined the assessment and identified possible modifications to the superstructures.
BYRON-UFSAR 3.7A-108 QUESTION 130.9 "The river screen house at the Byron station is founded on soil. Soil-structure interaction was performed by using the finite element method. It is the staff's position that the methods for implementing the soil-structure interaction analysis should include both the half space lumped spring and mass representation and the finite element approaches. Category I structures, systems and components should be designed to responses obtained by any one of the following methods: a. Envelope of results of the two methods, b. Results of one method with conservative design consideration of impact from use of the other method.
- c. Combination of a. and b. with provisions of adequate conservatism in design. "Therefore, we request you to compare the responses obtained by the half space (lumped parameter) approach to those obtained by the finite element approach at a few typical locations.
"Floor response spectra should be provided at least at the base mat, an intermediate elevation and an upper elevation. For the lumped parameter representation, the variation of soil properties should be considered." RESPONSE The soil-structure interaction responses are generated using the half-space lumped parameter approach. A soil shear modulus corresponding to a low value (10-4%) of strain is used in the analysis. The lower and upper bounds of the soil properties are considered (see Figure 2.5-89) of the Byron FSAR). The soil properties under the river screen house vary with depth. The average soil properties shown in Table Q130.9-1 were used in the elastic half-space soil spring analysis. The translational, rocking, and vertical soil-spring constants are computed based on these soil properties and formulas provided in Reference 1. The numerical values of soil spring constants are shown in Table Q130.9-2.
The horizontal floor response spectra at the base mat (el. 664 ft), an intermediate elevation (el. 702 ft), and the roof (el. 744 ft) are presented in Figures Q130.9-1 through Q130.9-6 for the OBE and in Figures Q130.9-7 through Q130.9-12 for the SSE. The response spectra are generated for 1% damping for OBE and 2% damping for SSE. The dot-dash line represents the design basis, using the finite element method (FEM). The solid line BYRON-UFSAR 3.7A-109 and dash line are those obtained by the half-space solution using the lower and upper bounds, respectively, of the soil properties. Table Q130.9-3 lists the comparison of the selected shear wall forces obtained using the soil spring soil-structure interaction (SSI) response spectra and the finite element SSI response spectra. Figures Q130.9-13 and Q130.9-14 provide the locations of the shear walls. The shear walls were evaluated to these higher forces and the stresses were found to be within the allowable.
The vertical soil-structure interaction analysis using the half-space approach has also been performed. The vertical floor response spectra at the base mat (elevation 664 ft), an intermediate elevation (elevation 702 ft), and the roof (elevation 744 ft) are presented in Figures Q130.9-15 through Q130.9-17 for the OBE and in Figures Q130.9-18 through Q130.9-20 for the SSE. The response spectra are generated for 1% damping for OBE and 2% damping for SSE. The dot-dash line represents the design basis, using the finite element method (FEM). The solid line and dash line are those obtained by the half-space solution using the lower and upper bounds, respectively, of the soil properties.
The higher responses, when the soil spring method is used, can be attributed directly to the conservative assumptions which are made in the analysis as compared to those used in the finite element method: (1) the river screen house is a deeply embedded structure, yet the soil spring method used required that the embedment be neglected; (2) the soil properties vary with depth, as shown in Figure 2.5-89 of the Byron FSAR: an average soil layer property was used in the soil spring method; (3) the soil-shear modulus corresponding to low shear strain and no material damping were used in the soil spring method, while more realistic strain-dependent shear modulus and damping values were used in the finite element method; and (4) in the present B/B seismic design criteria, Regulatory Guide 1.60 spectra are obtained at the surface and deconvolution analysis is used to obtain foundation elevation motions in the free field. These free field motions were used in the finite element SSI analysis; in the present soil spring analysis, the Regulatory Guide 1.60 spectra are applied at the foundation elevation.
In view of these assumptions, we feel that the soil spring SSI results are overly conservative. REFERENCES
- 1. "Analysis for Soil-Structure Interaction Effects for Nuclear Power Plants," Report by the Ad-Hoc Group on SSI of the Structural Division of ASCE, November 1978.
BYRON-UFSAR 3.7A-110 TABLE Q130.9-1 AVERAGE SOIL PROPERTIES AT THE RIVER SCREEN HOUSE SITE SHEAR MODULUS POISSON'S WEIGHT DENSITY (K/ft2) RATIO (kip/ft3) Lower Bound 3324 0.42 0.123 Upper Bound 4627 0.42 0.123
BYRON-UFSAR 3.7A-111 TABLE Q130.9-2 SPRING AND DASHPOT CONSTANTS NUMERICAL VALUES PARAMETER* UNITS (UPPER BOUND) (LOWER BOUND) Translational Stiffness, Kx Lb/in. 10.994 x 107 7.898 x 107 Translational Stiffness, Ky Lb/in. 10.944 x 107 7.898 x 107 Vertical Stiffness, Kz Lb/in. 14.350 x 107 10.310 x 107 Rocking Stiffness, xK Lb-in/rad 8.106 x 1013 5.823 x 1013 Rocking Stiffness, yK Lb-in/rad 3.127 x 1013 2.246 x 1013 Translational Damping, Cx Lb-sec/in. 3.259 x 106 2.763 x 106 Translational Damping, Cy Lb-sec/in. 3.259 x 106 2.763 x 106 Vertical Damping, Cz Lb-sec/in. 6.278 x 106 4.510 x 106 Rocking Damping, xC Lb-in-sec/rad 1.459 x 1012 1.263 x 1012 Rocking Damping, yC Lb-ins-sec/rad 3.935 x 1011 3.335 x 1011 ___________________
- The foundation shape and coordinate axes considered in the analysis are shown below:
BYRON-UFSAR 3.7A-112 TABLE Q130.9-3 COMPARISON OF SHEAR WALL FORCES FROM FINITE ELEMENT AND SOIL SPRING APPROACHES OBE SSE FEM* SSM** FEM SSM SHEAR WALL SPRING NO. (kips) (kips) (kips) (kips) X-1011 154 271 287 497 X-1012 198 356 391 670 X-1013 33 61 73 120 X-1014 33 61 73 120 X-1015 88 163 198 325 X-1016 28 53 65 106 X-1017 28 53 65 106 X-1018 46 86 108 175 Y-1021 9 16 22 35 Y-1022 9 16 22 35 Y-1023 102 185 250 403 Y-1024 62 114 155 247 Y-1025 63 115 155 250 Y-1026 168 311 424 670 Y-1027 170 314 424 680
- Finite element method ** Soil spring method BYRON-UFSAR 3.7A-113 QUESTION 130.9a* "For the river screen house at the Byron station there is a marked increase in the response spectra for most of the frequencies of interest in structural design. The technical position of the Regulatory staff is that the results of the two methods, i.e., the half space and the finite element method should be enveloped in order to be used in the design. This position is stated in the enclosure and designated as Method 3(a). As an alternate solution, the staff would find acceptable the two other options which are designated in the Enclosure as Methods 3(b) and 3(c). You are requested to perform a seismic analysis using one of the above noted three options, quantitatively assess its impact on structural design of the river screen house at the Byron plant and submit the results for our review." RESPONSE To comply with the staff position, all structures and components of the Byron river screen house will be qualified to the envelope of the responses based on the half-space and finite element methods for soil-structure interaction (SSI) analysis. Any modifications resulting from this reanalysis will be completed prior to plant operation. Consistent with the NRC staff position on Byron seismic reevaluation, requalification of the Byron river screen house structures using the enveloped spectra will be limited to the SSE load combinations only.
The details of the finite element method used for the SSI analysis are provided in Subsection 3.7.2.4. For the half-space approach, the Regulatory Guide 1.60 spectrum, normalized to 0.2g, was used as input to the soil-structure model. Frequency-dependent soil impedance functions were computed based on the method proposed by Luco (Reference 1). Two sets of soil properties, as presented in Table Q130.9a-1, were used. The soil properties in Set 1 are consistent with those used in the finite element method analysis. Set 2 corresponds to 40% of the geophysical soil shear modulus and represents an upper bound estimate of the soil shear modulus under the SSE excitation at the Byron site. The 40% value is an upper bound soil property of 10-3 inch/inch strain. The coupled soil structure system was analyzed in the frequency domain using the complex frequency response method.
- QUESTION 130.9a is a restated version of NRC QUESTION 130.9.
BYRON-UFSAR 3.7A-114 Modifications to the superstructure resulting from the reanalysis will consist of the following items as required: a. Vertical bracing will be added.
- b. Existing connections for the vertically braced column rows will be reinforced. c. Cover plates will be added to floor and roof beams. REFERENCES
- 1. J. E. Luco, "Vibrations of a Rigid Disc on a Layered Visoelastic Medium," from "Nuclear Engineering and Design," Vol. 36, pp. 325-340, 1976.
BYRON-UFSAR 3.7A-115 TABLE Q130.9a-1 SSE SOIL PROPERTIES USED FOR HALF-SPACE SSI ANALYSIS LAYER DEPTH TO TOP LAYER SHEAR MODULUS (ksf) NUMBER OF LAYER (ft) THICKNESS (ft) SET 1 SET 2 1 0 16 399.0 1850.0 2 16 18 760.0 1850.0 3 34 24 1219.0 1850.0 4 58 32 1997.0 1850.0 5 90 half-space 45000.0 45000.0
B/B-UFSAR 3.7A-116 3.7A.6 Braidwood Site Specific Seismic Concerns NRC Question 362.11 challenged the adequacy of the seismic design basis for all Category I structures not founded on rock. The response noted that although the lake screen house rested on 10 feet of hard glacial till there was no appreciable amplification between the rock and the top of the till. NRC Question 130.56 requested a comparison of the half-space lumped parameter method with the finite element method for determining responses in the lake screen house. The response provided the comparison via reference to Question 130.6a contained in Section 2 of this attachment.
BRAIDWOOD-UFSAR 3.7A-117 QUESTION 362.11 "The safe shutdown earthquake for the Braidwood Station site is based on the postulated occurrence of a maximum Modified Mercalli intensity VIII (body wave magnitude about 5.8) earthquake near the site (see Byron Station Safety Evaluation Report (SER), NUREG-0876). The staff's position, as stated in the Byron Station SER, is that a Regulatory Guide 1.60 spectrum with a high frequency anchor of 0.20g at the foundation level of structures founded on rock is an adequately conservative representation of the vibratory ground motion from this size earthquake. Soils can amplify vibratory ground motion. The amplitude and frequency of the amplified motion is a function of the physical properties of the material and its thickness. For all Category I structures not founded on rock, demonstrate the adequacy of the design basis by directly calculating a site-specific response spectrum and/or by calculating the amplification of an appropriate rock spectrum resulting from the presence of the soil. "We recommend that the details of the study planned in response to this question be discussed with the staff prior to work initiation." RESPONSE For the Braidwood site, the design response spectra, which is defined at the ground, is a 0.26g Regulatory Guide 1.60 spectra as shown in Figure 2.5-47. Foundation level response spectra and time histories were generated by deconvolution as described in Subsection 3.7.1. The maximum horizontal and vertical ground accelerations at the foundation level is 0.2g for SSE.
In response to Question 130.6a, all structures and equipment required for cold shutdown were reevaluated for a 0.2g Regulatory Guide 1.60 spectra specified at the foundation elevation of all Category I structures. The founding conditions of the various Category I structures and justification for the use of 0.2g Regulatory Guide 1.60 spectra at the foundation elevation is as follows.
Lake Screen House The lake screen house rests on 10 feet of hard glacial till. The shear wave velocity of the till material is 2,400 ft/sec. The shear wave velocity of the underlying rock is 3,200 ft/sec. Because of the high soil column frequency (Vs/4H=60 Hz) and the low velocity contrast (3,200/2,400=1.33) between the rock and the soil medium, there will be no appreciable amplification of motion between the rock and the top of the till in the critical frequency range of 1 to 20 Hz. Thus, the
BRAIDWOOD-UFSAR 3.7A-118 use of the 0.2 g Regulatory Guide 1.60 spectrum at the foundation elevation of the lake screen house for reevaluation in response to Question 130.6a is justified. Containment Structure The containment structure is founded on rock.
Auxiliary Building - Fuel Handling Building Complex The auxiliary-fuel handling building complex is founded on rock with only a small percentage of the foundation at the periphery overlying soil (see Figures 2.5-16 and 3.8-45). The major slabs of the structure are continuous diaphragms and monolithically connect the portions on soil to the portions on the rock foundation. Thus, the structure is essentially founded on rock and was analyzed as such.
BRAIDWOOD-UFSAR 3.7A-119 QUESTION 130.56 "Conflicting information is provided in Table 3.7-3 of Section 3.7 and Section 3.8.5.1.5 of the FSAR which state that the Braidwood lake screen house rests on rock and glacial till respectively. If the information contained in Section 3.8.5.1.5 is correct, describe the method of analysis used to account for the soil-structure interaction. "It is the staff's position that the methods implementing the soil-structure interaction analysis for structure situated on soil should include both the half space lumped spring and mass representation and the finite element approaches. Category I structures, systems and components should be designed to responses obtained by the appropriate methods described in the Standard Review Plan, Section 3.7.2.II.4. "Therefore, you are requested to compare the responses obtained by the half space (lumped parameter) method to those obtained by the finite element approach at the following locations: "(a) Base mat (b) El. 588'-0" (c) El. 602'10" (d) Roof. "In both analyses the variation of soil properties should be considered. "If the lake screen house is supported by bedrock, please confirm that the information provided in Section 3.8.5.1.5 on the same subject is incorrect and make the necessary FSAR corrections."
RESPONSE The Braidwood lake screen house foundation rests on 10 feet of glacial till underlain by rock. Due to the shallow depth (10 feet) of the till and a high shear wave velocity of 2400 ft/sec, the soil structure interaction effects are not significant (see the response to Question 362.11). The original design for structures and components was based on a fixed base analysis using the deconvoluted rock time history; the comparison between the design spectra and the spectra from the deconvoluted rock time history is shown in Figures 3.7-21 through 3.7-40. Furthermore, as a result of Question 130.6a, a reanalysis was performed on a fixed base analysis using the design time BRAIDWOOD-UFSAR 3.7A-120 history. The main slabs and walls and the major equipment are designed for the envelope of the two analyses results. Table 3.7-3 has been revised to indicate there is 10 feet of hard glacial till between the rock and the lake screen house foundation.
B/B-UFSAR 3.8-67 REVISION 9 - DECEMBER 2002 TABLE 3.8-1 CONTAINMENT PENETRATIONS VERTICAL HORIZONTALPENETRATIONWALLELEVATIONSKEWSKEWNUMBERSIZE THICKNESS (ft-in.)AZIMUTH (degree-min)ANGLEANGLEDESCRIPITION B/B-UFSAR 3.8-68 REVISION 2 - DECEMBER 1990 TABLE 3.8-1 (Cont'd) VERTICAL HORIZONTALPENETRATIONWALLELEVATIONSKEWSKEWNUMBERSIZE THICKNESS (ft-in.)AZIMUTH (degree-min)ANGLEANGLEDESCRIPITION B/B-UFSAR 3.8-69 REVISION 2 - DECEMBER 1990 TABLE 3.8-1 (Cont'd) VERTICAL HORIZONTALPENETRATIONWALLELEVATIONSKEWSKEWNUMBERSIZE THICKNESS (ft-in.)AZIMUTH (degree-min)ANGLEANGLEDESCRIPITION B/B-UFSAR 3.8-70 REVISION 7 - DECEMBER 1998 TABLE 3.8-1 (Cont'd) VERTICAL HORIZONTALPENETRATIONWALLELEVATIONSKEWSKEWNUMBERSIZE THICKNESS (ft-in.)AZIMUTH (degree-min)ANGLEANGLEDESCRIPITION B/B-UFSAR 3.8-71TABLE 3.8-1 (Cont'd) VERTICAL HORIZONTALPENETRATIONWALLELEVATIONSKEWSKEWNUMBERSIZE THICKNESS (ft-in.)AZIMUTH (degree-min)ANGLEANGLEDESCRIPITION B/B-UFSAR 3.8-72 REVISION 3 - DECEMBER 1991 TABLE 3.8-1 (Cont'd) VERTICAL HORIZONTALPENETRATIONWALLELEVATIONSKEWSKEWNUMBERSIZE THICKNESS (ft-in.)AZIMUTH (degree-min)ANGLEANGLEDESCRIPITION B/B-UFSAR 3.8-73 REVISION 1 - DECEMBER 1989 TABLE 3.8-2 LIST OF SPECIFICATIONS, CODES, AND STANDARDS* SPECIFICATION SPECIFICATIONREFERENCEOR STANDARDNUMBERDESIGNATIONTITLE1ACI 318-71,77,83 Building Code Requirements for Reinforced Concrete 2ACI 301Specifications for Structural Concrete for Buildings 3 ACI 347 Recommended Practice for ANSI A145.1 Concrete Formwork 4ACI 305Recommended Practice for Hot ANSI A170.1Weather Concreting 5 ACI 211.1 Recommended Practice for Selecting Proportions for Normal Weight Concrete 6 ACI 304 Recommended Practice for Measuring, Mixing, Trans-porting, and placing concrete 7 ACI 315 Manual of Standard Practice for Detailing Reinforced Concrete Structures 8 ACI 306 Recommended Practice for Cold Weather Concreting 9 ACI 309 Recommended Practice for Consolidation of Concrete 10ACI 308Recommended Practice for Curing Concrete 11 ACI 214 Recommended Practice for ANSI A146.1 Evaluation of Compression Test Results of Field *References to edition dates are shown for the codes used inthe design of safety-related structures. All otherspecifications delineated in this table are recommendedpractices and material specifications that do not affect thedesign of safety-related structures.
B/B-UFSAR 3.8-74 REVISION 7 - DECEMBER 1998 TABLE 3.8-2 (Cont'd) ** Clarifications to, and deviations from portions of AWS D1.1, "Structural Welding Code," are made based on engineering" evaluations. Visual weld inspection requirements are based on guidelines in a document prepared by the Nuclear Construction Issues Group, NCIG-01, Revision 2, "Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants."SPECIFICATION SPECIFICATIONREFERENCEOR STANDARDNUMBERDESIGNATIONTITLE12 ACI 311 Recommended Practice for Concrete Inspection 13ACI 304Preplaced Aggregate Concrete for Structural and Mass Concrete 14Report by ACI Placing Concrete by Pumping Committee 304 Method15AISC-69,78 Specification for the Design, Fabrication, and Erection of Structural Steel for Building 16AWS D1.1** Structural Welding Code 17 ASME Boiler & Pressure Vessel Code,Section III ASME-1971, S73 Division 1, Subsection NE ASME-1974, S75 Division 1, Subsection NF ASME-1973 Division 2, Proposed Standard Code for Concrete Reactor Vessels and Containments Issued for Trial Use and Comments ASME-1980 Division 2, CC 6000 ASME-1992 1992 Addenda, Division 1,Section XI, Subsection IWL, IWE 18 American PublicTest Methods Sulphides in Health Assoc. Water, Standard Methods for (APHA)the Examination of Water and Waste Water 19 ASTM Annual Books of ASTM Standards20CRSIManual of StandardMSP-1Practice B/B-UFSAR 3.8-75 REVISION 1 - DECEMBER 1998 TABLE 3.8-2 (Cont'd) SPECIFICATION SPECIFICATION REFERENCE OR STANDARD NUMBER DESIGNATION TITLE 21 ANSI N45.2.5 Proposed Supplementary Q.A. Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During Construction Phase of Nuclear Power Plants 22 CRD Chief of Research and Development Standards, Department of the Army, Handbook for Concrete and Cement Volume I and II, Corps of Engineers U.S. Army 23 ACI-349-76,85 Code Requirements for Nuclear Safety Related Concrete Structures 24 AISI Specification for design of cold-formed steel structural members EXPLANATION OF ABBREVIATIONS ACI - American Concrete Institute
AISC - American Institute of Steel Construction AISI - American Iron and Steel Institute ANSI - American National Standards Institute ASME - American Society of Mechanical Engineers ASTM - American Society of Testing Materials AWS - American Welding Society
CRD - Chief of Research and Development Standards CRSI - Concrete Reinforcing Steel Institute ___________________ NOTE: For exceptions and additions to these codes and standards refer to Appendix B.
B/B-UFSAR 3.8-76 TABLE 3.8-3 LOAD COMBINATIONS AND LOAD FACTORS CONTAINMENT-SERVICE LOAD AND FACTORED LOAD CONDITIONS LOAD COMBINATION CATEGORY (e) LOAD FACTORS (d) (c) D L F P' Pa Po To Ta E E' W W' Ro Ra Yr Ym Yj H' M I CONSTRUCTION 1 0.75 0.75 1.0 0.75 0.75(a) 2 1.0 1.0 1.0 1.0
II TEST 3 1.0 1.0 1.0 1.0 1.0 III NORMAL 4 1.0 1.0 1.0 1.0 1.0 1.0 IV SEVERE 5 1.0 1.0 1.0 1.0 1.0 1.0 1.0 ENVIRONMENTAL 6 1.0 1.0 1.0 1.0 1.0 1.0 1.0
V ABNORMAL 7 1.0 1.0 1.0 1.5 1.0(b) 1.0 1.0 8 1.0 1.0 1.0 1.0 1.0
VI EXTREME 9 1.0 1.0 1.0 1.0 1.0 1.0 1.0 ENVIRONMENTAL
10 1.0 1.0 1.0 1.0 1.0 1.0 1.0
11 1.0 1.0 1.0 1.0 1.0 1.0 1.0 VII ABNORMAL/ 12 1.0 1.0 1.0 1.25 1.0(b) 1.25 1.0 1.0 1.0 1.0 SEVERE ENVIRONMENTAL 13 1.0 1.0 1.0 1.25 1.0(b) 1.25 1.0 1.0 1.0 1.0
B/B-UFSAR 3.8-77 TABLE 3.8-3 (Cont'd) LOAD COMBINATIONS AND LOAD FACTORS CONTAINMENT-SERVICE LOAD AND FACTORED LOAD CONDITIONS LOAD COMBINATION CATEGORY (e) LOAD FACTORS (d) (c) D LFP' Pa Po To TaE E' WW' Ro Ra Yr Ym Yj H' M VIII ABNORMAL/ 14 1.0 1.0 1.01.01.0(b) 1.01.0 1.0 1.0 1.0 EXTREMEENVIRONMENTAL___________________________ NOTES: (a) For the construction category, the wind load for a ten year recurrence interval is used. (b) Ta is based on a temperature corresponding to the factored pressure. (c) Loads not applicable to a particular system are deleted. (d) For any load combination, loads other than D are deleted when necessary to produce the most severe combination of loads.
(e) Categories I through IV present information for severe load designs; Categories V through VIII present information for fractured loads.
B/B-UFSAR 3.8-78 REVISION 4 - DECEMBER 1992 TABLE 3.8-4 DEFINITIONS OF STRUCTURAL TERMINOLOGY DEFINITIONS NOTE: These definitions apply to the terms used in Tables 3.8-7 through 3.8-11. I LOADING CATEGORIES CONSTRUCTION All events and loads during structural construction including the various stages of prestressing, but excluding those during testing. TESTING All events and loads applied during structural integrity tests and preoperational tests such as hydrostatic testing of equipment and the pressure tests. Each testing event is considered to be mutually exclusive of other testing events. NORMAL All events and loads that could reasonably be expected during the operation, shutdown, and normal maintenance of the power plant. SEVERE ENVIRONMENTAL All loads due to infrequent site-related environmental events like operating-basis earthquake and design wind. ABNORMAL All loads due to postulated accident events. They include pressure, temperature, pipe whip, jet impingement, and pipe reactions due to each break postulated for the design-basis accidents. This loading condition also includes plant-related nonenvironmental missiles. The loads from each postulated accident event are considered to be mutually exclusive of other postulated accidents.
B/B-UFSAR 3.8-79 TABLE 3.8-4 (Cont'd) EXTREME ENVIRONMENTAL All loads due to site-related environmental events which are credible but highly improbable. These events include the safe shutdown earthquake, design-basis tornado, probable maximum flood, and the postulated site-related accidents not included in the abnormal loading category. ABNORMAL/SEVERE Loads due to the highly improbable simultaneous occurrence ENVIRONMENTAL of abnormal and severe environmental loading categories. Only the specified combinations of these categories are considered. ABNORMAL/EXTREME Loads due to the extremely improbable simultaneous ENVIRONMENTAL occurrence of the abnormal and extreme environmental loading conditions. Only the specified combinations of these conditions are considered. II LOADS DEAD D - The dead load of structure plus any other permanent load (except prestressing forces), including vertical and lateral pressures of liquids, piping, cable pans, weight of permanent equipment and its normal contents under operating conditions and the weight of soil cover for buried structures. LIVE L - Conventional floor and roof live loads, movable equipment loads, crane loads and other loads which vary in intensity and occurrence such as lateral soil pressure. The dynamic effect of operating equipment is accounted for by the use of appropriate impact factors. PRESTRESS F - Loads resulting from the application of prestressing forces.
OPERATING PRESSURE Po - The normal operating pressure differentials.
B/B-UFSAR 3.8-80 REVISION 3 - DECEMBER 1991 TABLE 3.8-4 (Cont'd) ACCIDENT PRESSURE Pa - The design accident pressure load. This pressure is based upon peak calculated pressure with appropriate margin provided for uncertainties in this calculation: Containment internal pressure, + 50.0 psig. Containment external pressure, +/- 3.0 psig. TEST PRESSURE P' - The containment's test pressure, + 57.5 psig. OPERATING PIPE Ro - Normal operating or shutdown pipe reactions at REACTION supports or anchor points, based on the most critical transient or steady-state condition. ACCIDENT PIPE Ra - Pipe reactions generated by the design-basis accident REACTIONincludingRo. THERMALTo - Thermal effects and loads during normal operating, shutdown, construction and test conditions, including average temperature and temperature gradients. The combination of internal and ambient temperatures which produces the most critical transient or steady-state thermal gradient is used. Climatic temperature ranges: Maximum outside temperature: 110° F Minimum outside temperature: -25.0° F. ACCIDENT THERMAL Ta - Thermal effects and loads generated by the design-basis accident including To.
B/B-UFSAR 3.8-81 REVISION 4 - DECEMBER 1992 TABLE 3.8-4 (Cont'd) OPERATING BASIS E - Loads generated by the operating-basis earthquake EARTHQUAKE (OBE), including dynamic lateral soil pressure and hydrodynamic groundwater pressure for a horizontal ground acceleration at foundation elevation of 0.09g. SAFE SHUTDOWN E' - Loads generated by the safe shutdown earthquake (SSE), EARTHQUAKE including dynamic lateral soil pressure and hydrodynamic groundwater pressure for a horizontal ground acceleration at foundation elevation of 0.20g.
WIND W - Design Wind Load. TORNADO Wt - Tornado loads including the effects of wind pressure, differential pressure loads due to rapid atmospheric pressure change and missile impact. MAXIMUM PROBABLE H' - Forces associated with the maximum probable flood, FLOOD seiche, or precipitation.
MISSILE M - Loads associated with missiles other than a tornado (see Section 3.5). RESTRAINT Yr - Equivalent static load on the restraint generated by the reaction of broken high-energy pipe during the postulated break, and including an appropriate dynamic factor to account for the dynamic nature of the load. JET IMPINGEMENT Yj - Jet impingement equivalent static load on a structure generated by the postulated break, and including an appropriate dynamic factor to account for the dynamic nature of the load.
PIPE WHIP MISSILE Ym - Missile impact equivalent static load on a structure generated by or during the postulated break, as from pipe whipping, and including an appropriate dynamic factor to account for the dynamic nature of the load.
B/B-UFSAR 3.8-82 REVISION 4 - DECEMBER 1992 TABLE 3.8-4 (Cont'd) EXTERNAL PRESSURE Pe - External pressure on containment, not considering Pa. BREAK LOAD Rr - Load associated with a high energy break of a piping system = Ym + Yj + Yr.
B/B-UFSAR 3.8-83 TABLE 3.8-5 ALLOWABLE STRESSES AND STRAINS CONTAINMENT LINER PLATE AND ANCHORAGES LINER PLATE ALLOWABLES STRESS/STRAIN ALLOWABLE* The allowables are defined as: fst = allowable liner plate tensile stress, psi; fsc = allowable liner plate compressive stress, psi; sc = allowable liner plate compressive strain; and st = allowable liner plate tensile strain. *The types of strains limited by this table are strains induced by deformation or constraint. COMBINED MEMBRANE CATEGORY MEMBRANE AND BENDING Construction fst = fsc = 2/3Fy fst = fsc = 2/3Fy Service sc = 0.002 in./in. sc = 0.004 in./in.
st = 0.001 in./in. st = 0.002 in./in. Factored sc = 0.005 in./in. sc = 0.014 in./in.
st = 0.003 in./in. st = 0.010 in./in.
B/B-UFSAR 3.8-84 TABLE 3.8-5 (Cont'd) LINER ANCHOR ALLOWABLES FORCE/DISPLACEMENT ALLOWABLES DISPLACEMENTCATEGORYMECHANICAL LOADS LIMITED LOADS TestLesser of a = 0.25 u Fa = 0.67FyFa = 0.33FuNormalLesser of: a = 0.25 u Fa = 0.67FyFa = 0.33FuSevere EnvironmentalLesserof a = 0.25 u Fa = 0.67FyFa = 0.33FuExtreme EnvironmentalLesserof a = 0.25 u Fa = 0.67FyFa = 0.33FuAbnormalFa = 0.9Fy a = 0.50 u Fa = 0.5FuAbnormal/severe environmental Fa = 0.9Fy a = 0.50 u Fa = 0.5FuAbnormal/extreme environmental Fa = 0.9Fy a = 0.50 u Fa = 0.5Fu B/B-UFSAR 3.8-85 REVISION 3 - DECEMBER 1991 TABLE 3.8-5 (Cont'd) LINERPLATEANCHORAGES(stress/strains)(force/displacement allowables) COMBINED MEMBRANEMECHANICAL DISPLACEMENTCATEGORYMEMBRANEAND BENDINGLOADSLIMITED LOADSConstructionfst = fsc = 0.67Fy fst = fsc = 0.67Fy LesserofTest sc= 0.002 in/insc= 0.004 in/in Fa = 0.67Fy a= 0.25 uFa = 0.33 Fust= 0.001 in/inst= 0.002 in/in Normal sc= 0.002 in/insc= 0.004 in/in st= 0.001 in/inst= 0.002 in/in Severe Environmental sc= 0.002 in/insc= 0.004 in/in st= 0.001 in/inst= 0.002 in/in LesserofAbnormal sc= 0.005 in/insc= 0.014 in/in Fa = 0.90Fy a= 0.5 uFa = 0.50 Fust= 0.003 in/inst= 0.010 in/in Extreme Environmental sc= 0.002 in/insc= 0.004 in/in st= 0.001 in/inst= 0.002 in/in Abnormal Severe sc= 0.005 in/insc= 0.014 in/in st= 0.003 in/inst= 0.010 in/in Abnormal Extreme sc= 0.005 in/insc= 0.014 in/in st= 0.003 in/inst= 0.010 in/in B/B-UFSAR 3.8-86 TABLE 3.8-6 PREDICTED CONTAINMENT DEFLECTIONS DURING THE STRUCTURAL ACCEPTANCE TEST I. RADIAL DEFLECTION OF THE CYLINDRICAL WALL OUTWARDMETER NUMBER LOCATION DEFLECTION II.VERTICAL GROWTH WITH REFERENCE TO THE FOOTOF THE CYLINDRICAL WALL UPWARDVERTICALMETER NUMBER LOCATION GROWTH (inch)
B/B-UFSAR 3.8-87 TABLE 3.8-6 (Cont'd) UPWARDVERTICALMETER NUMBER LOCATION GROWTH (inch)III.TANGENTIAL DEFLECTION OF EQUIPMENT HATCH OPENINGTANGENTIALMETERDEFLECTIONNUMBERLOCATION(inch)
B/B-UFSAR 3.8-88 TABLE 3.8-7 LOAD DEFINITIONS AND COMBINATIONS FOR CLASS MC CONTAINMENT COMPONENTS* (For Definitions See Table 3.8-4) LOAD FACTORS LOADINGITEMSEVEREEXTREMECATEGORYNUMBERENVIRONMENTALABNORMALENVIRONMENTALD L RoTo Po Pe P' E Rr Ta Pa Ra M E' Construction11.0 1.01.0Test2 1.0 1.0 1.0** 1.0Normal31.0 1.0 1.01.0 1.0 1.0SevereEnvironmental41.0 1.0 1.01.0 1.0 1.01.0Abnormal51.0 1.01.0 1.0 1.0 1.061.0 1.01.01.0 1.0 1.0 1.0ExtremeEnvironmental71.0 1.0 1.0 1.0 1.01.0*Does not include process piping penetrations. **Temperature at time of test.
B/B-UFSAR 3.8-89 TABLE 3.8-7 (Cont'd) LOAD FACTORS LOADING ITEM SEVERE EXTREME CATEGORY NUMBER ENVIRONMENTAL ABNORMAL ENVIRONMENTAL D L Ro To Po Pe P' E Rr Ta Pa Ra M E' Abnormal/Severe Environmental 8 1.0 1.0 1.0 1.0 1.0 1.0 1.0 9 1.0 1.0 1.0 1.0 1.0 1.0 10 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Abnormal/Extreme Environmental 11 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 12 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0
B/B-UFSAR 3.8-90 TABLE 3.8-8 PHYSICAL PROPERTIES FOR MATERIALS TO BE USED FOR PRESSURE PARTS OR ATTACHMENT TO PRESSURE PARTS MC COMPONENTS Sm Sy Sy ASME CODE Su MINIMUM MINIMUM ALLOWABLE MINIMUM YIELD YIELD STRESS ULTIMATE AT THE AT INTENSITY MATERIAL TENSILE AMBIENCE340°F AT 340°f SPECIFICATION (ksi) (ksi) (ksi) (ksi) NOTES Plate SA516 Gr 70 70 38 33.26 17.5 SA516 Gr 60 60 32 27.94 15 SA240 Tp 304 75 30 21.78 16.44
Pipe SA106 Gr B 60 35 30.6 15 SA333 Gr 6 60 35 30.6 15 SA333 Gr 1 55 30 26.24 13.75 SA312 Type 304 75 30 21.78 16.44 SA376 Type 304 75 30 21.78 16.44 Forgings and Fittings SA350 LF-1 60 30 26.24 15.0 SA350 LF-2 70 36 31.96 17.5 SA182 F304 70 30 21.78 16.44 SA182 Gr-F316 75 30 18.22 18.28 Bolting SA193 B7 115 95 83.98 23 Between 2.5 in. and 4 in.
diameter SA193 B7 125 105 93.06 25 Under 2.5 in. diameter
B/B-UFSAR 3.8-91 TABLE 3.8-8 (Cont'd) Sm Sy Sy ASME CODE Su MINIMUM MINIMUMALLOWABLE MINIMUM YIELD YIELD STRESS ULTIMATE AT THE AT INTENSITY MATERIAL TENSILE AMBIENCE 340°F AT 340°f SPECIFICATION (ksi) (ksi) (ksi) (ksi) NOTES SA194 Gr 7* --- --- --- ---
SA320 L43 125 1-5 94.14 25 4 in. diameter and under
_________________________
- No yield or tensile strength specified. Assume it is same as an equivalent grade in SA-193-B7.
B/B-UFSAR 3.8-92 REVISION 10 - DECEMBER 2004 TABLE 3.8-9 CATEGORY I LOAD COMBINATION TABLE STRUCTURAL STEEL ELASTIC DESIGN LOAD FACTORS LOADING CONDITIONS D L S Ro Ra E E' W Wt To Ta Pa H' M Yr Yj Ym CONSTRUCTION 1 1.0 1.0 1.0 1.0 TEST 2 1.0 1.0 1.0 1.0 1.0 NORMAL 3 1.0 1.0 1.0 1.0 1.0 SEVERE 4 1.0 1.0 1.0 1.0 1.0 ENVIRONMENTAL 5 1.0 1.0 1.0 1.0 1.0 ABNORMAL 6 1.0 1.0 1.0 1.0 1.0 1.0 7 1.0 1.0 1.0 1.0 1.0 1.0 EXTREME 8 1.0 1.0 1.0 1.0 1.0 ENVIRONMENTAL 9(6) 1.0 1.0 1.0 1.0 1.0 10(6) 1.0 1.0 1.0 1.0 1.0
B/B-UFSAR 3.8-93 REVISION 10 - DECEMBER 2004 TABLE 3.8-9 (Cont'd) LOAD FACTORS LOADING CONDITIONS D L S Ro Ra E E' W Wt To Ta Pa H' M Yr Yj Ym ABNORMAL/SEVERE 11 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 ENVIRONMENTAL
ABNORMAL/EXTREME 12 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 ENVIRONMENTAL
_______________________________
NOTES: (1) The construction loading combination is used to check the capacity of structural members for construction loading. The live load used is specified by the contractor to reflect shoring loads, etc. For the construction category, the wind load corresponding to the ten year recurrence interval is used. (2) For any load combination, loads other than D are deleted when necessary to produce the most severe combination of loads. (3) Loads not applicable to a particular system are deleted.
(4) For both E and E', the resultant effects (resultant stresses) for both horizontal and vertical earthquake forces are determined by combining the individual effects by the square root of the sum of the squares method.
(5) AISC allowables are calculated in accordance with Part I of AISC - 69.
(6) Loading conditions number 9 and 10 are not applicable to Byron River Screen House.
B/B-UFSAR 3.8-94 REVISION 10 - DECEMBER 2004 TABLE 3.8-10 CATEGORY I LOAD COMBINATION TABLE REINFORCED CONCRETE ULTIMATE STRENGTH DESIGN LOAD FACTORS LOAD CONDITIONS D L Ro Ra E E' W Wt To Ta Pa H' M Yr Yj Ym CONSTRUCTION 1 1.1 1.3 1.3 1.3 TEST 2 1.1 1.3 1.3 1.3
NORMAL 3 1.4 1.7 1.3 1.3
SEVERE ENVIRONMENTAL 4 5 1.4 1.2 1.7 1.3 1.3 1.7 1.7 1.3 1.3 6 1.4 1.7 1.3 1.9 1.3 7
ABNORMAL 8 1.0 1.0 1.0 1.0 1.5
9 1.0 1.0 1.0 1.0 1.0
EXTREME
10
1.0
1.0
1.0
1.0 1.0 ENVIRONMENTAL 11 1.0 1.0 1.0 1.0 1.0
12 1.0 1.0 1.0 1.0 1.0
ABNORMAL/SEVERE 13 1.0 1.0 1.0 1.25 1.0 1.25 1.0 1.0 1.0 ENVIRONMENTAL
B/B-UFSAR 3.8-95 REVISION 10 - DECEMBER 2004 TABLE 3.8-10 (Cont'd) LOAD FACTORSLOAD CONDITIONS D L RoRaE E' W Wt To Ta Pa H' M Yr Yj Ym ABNORMAL/EXTREME 14 1.0 1.01.01.01.0 1.01.0 1.0 1.0ENVIRONMENTAL__________________________ NOTES: (1) The construction loading combination is used to check the capacity of structural members for construction loading. The live load used is specified by the contractor to reflect shoring loads, etc. For construction category, the wind load for a ten year recurrence interval is used. (2) For any load combination, loads other than D are deleted when necessary to produce the most severe combination of loads. (3) Loads not applicable to a particular system are deleted. (4) For both E and E', the resultant effects for both horizontal and vertical earthquake are determined by combining the individual effects by the square root of the sum of the squares.
BYRON-UFSAR 3.8-96 REVISION 10 - DECEMBER 2004 TABLE 3.8-11 CATEGORY I LOAD COMBINATION TABLE FOR RIVER SCREEN HOUSE (BYRON) REINFORCED CONCRETE ULTIMATE STRENGTH DESIGN LOAD FACTORS LOAD CONDITION D L Ro E E' W To Ta H" H"' CONSTRUCTION 1 1.1 1.3 1.3 1.3 NORMAL 2 1.4 1.7 1.3 1.3
SEVERE 3
1.4 1.7 1.3 1.7 1.3 ENVIRONMENTAL 4 1.2 1.3 1.7 1.3 5 1.4 1.7 1.3 1.9 1.3 1.4 6 1.2 1.3 1.9 1.3 1.2
BYRON-UFSAR 3.8-97 REVISION 10 - DECEMBER 2004 TABLE 3-8.11 (Cont'd) LOAD FACTORS LOAD CONDITION D L Ro E E' W To Ta H" H"' EXTREME 7 1.0 1.0 1.0 1.0 1.0 1.0 ENVIRONMENTAL 8 1.0 1.0 1.0 1.0 1.0
9 1.0 1.0 1.0 1.0 1.0 1.0
__________________________
NOTES: (1) The construction loading combination is used to check the capacity of structural members for construction loading. The live load used is specified by the Contractor to reflect shoring loads, etc. For construction category, the wind load for a ten year recurrence interval is used. (2) For any load combination, loads other than D are deleted when necessary to produce the most severe combination of loads. (3) Loads not applicable to a particular system are deleted. (4) For both E and E' the resultant effects for both horizontal and vertical earthquake are determined by combining the individual effects by the square root of the sum of the squares. (5) H" refers to maximum flood of record.
(6) H"' refers to combined event flood.
B/B-UFSAR 3.8-98 REVISION 4 - DECEMBER 1992 TABLE 3.8-12 LOADING COMBINATIONS FOR PENETRATION SLEEVES AND HEAD FITTINGS LOADING PRIMARY +
LOADING COMPONENTS CONDITIONS DESIGN EXPANSION SECONDARY EMERGENCY FAULTED FATIGUE TESTING Internal and External Maximum Operating X X X Pressure Design X Transient X X Test X Temperatures Normal Operating X X X X X Design X Transient X X Weight X X X X Thermal Expansion X X X Seismic OBE X X X SSE X1 Relative Displacement X X X Fluid Dynamics Hydraulic Transients X X X X Main Steam SRV X X X Pipe Break and Jet X Impingement __________________________
Note 1 - SSE is a faulted condition load. However, the type of analysis used (system inelastic and component elastic) does not require the SSE load as input.
B/B-UFSAR 3.8-99 TABLE 3.8-13 ALLOWABLE STRESSES FOR PENETRATION SLEEVES AND HEAD FITTINGS ALLOWABLE STRESS VALUES FOR EACH LOADING CONDITION (Note 1) STRESS NORMAL DESIGN EMERGENCY FAULTED CATEGORY AND UPSET (Note 3) (Note 3) (Notes 3 and 4) PRIMARY STRESSES GENERAL (Note 2) Sm The larger The larger of MEMBRANE of 1.2Sm, 0.7Su, or (Pm) or Sy 3SSSyuy+ LOCAL (Note 2) 1.5Sm The larger The larger of MEMBRANE of 1.8Sm, 1.05Su, or (PL) or 1.5Sy 2SSS5.1YuY+ MEMBRANE (Note 2) 1.5Sm The larger The larger of + BENDING of 1.8Sm, 1.05Su, or (PL+PB) or 1.5Sy 2SSS5.1YuY+
B/B-UFSAR 3.8-100 TABLE 3.8-13 (Cont'd) STRESS NORMAL DESIGN EMERGENCY FAULTED CATEGORY AND UPSET (Note 3) (Note 3) (Notes 3 and 4) SECONDARY STRESSES EXPANSION 3Sm STRESSES (Pe) PRIMARY + 3Sm SECONDARY (PL+PB+Pe+Q) PEAK STRESSES (F) (Note 5) NOTES 1. Values for Sm, Sy, and Su shall be temperature-dependent and taken from Section III Tables, as follows: Sy from Tables I-2.0; Su from Tables I-3.0; Sm from Tables I-1.0 for non-MC components, and from Tables I-10.0 for MC components. 2. There are no specific limits established on the primary stresses that result from Operating Conditions. 3. Design, emergency and faulted conditions do not require Secondary and Peak stress evaluation. 4. The specified stress limits for faulted conditions are applicable for system inelastic and component elastic evaluation.
- 5. Used in combination with all primary and secondary stresses for calculating alternating stresses (for fatigue evaluation).
B/B-UFSAR 3.8-101 REVISION 1 - DECEMBER 1989 TABLE 3.8-14 COMPARISON OF THIN SHELL VS THICK SHELL ANALYTICAL RESULTS* RADIUS HEIGHT MERIDIONAL FORCE K MERIDIONAL MOMENT K-Ft. HOOP FORCE K HOOP MOMENT K-(ft) SECTION R(ft) Z(ft) THIN THICK VAR. % THIN THICK VAR. % THIN THICK VAR. % THIN THICK VAR. % 1 71.8 20.0 259.0 244.0 6.2 92.60 78.40 18.1 428.0 400.0 7.0 15.70 8.2 9.1 2 71.8 100.0 259.0 246.0 5.3 0.17 0.00 -- 504.0 503.0 0.2 0.03 6.2 -- 3 71.8 182.0 260.0 247.0 5.3 120.00 77.30 5.5 147.0 149.2 1.5 21.00 15.2 38.2 4 71.8 187.5 259.0 248.0 4.4 324.00 261.00 24.1 67.6 52.2 29.5 56.10 49.0 14.5 5 70.4 200.0 242.0 254.0 4.7 1324.00 1208.00 9.6 34.0 68.5 50.4 106.00 148.9 28.8 6 65.5 202.0 245.0 269.0 8.9 594.0 609.00 2.5 15.0 32.3 53.6 60.10 74.3 19.1 7 59.5 206.0 309.0 301.0 2.7 290.00 270.00 7.4 77.0 37.2 107.0 14.90 13.7 8.9 8 35.0 220.0 387.0 430.0 10.0 29.30 30.00 2.3 312.0 390.0 20.0 24.60 27.0 8.9
_________________________
- Variation is recorded as % of thick shell stress resultants.
Thin shell definition: Wall thickness < 0.1 (Radius of Curvature)
(
Reference:
W. Flugge, "Stresses in Shells" Springer-Verlag, 1960)
B/B-UFSAR 3.8-102 TABLE 3.8-15 COMPARISON OF ALLOWABLE STRESSES IN PSI (INSPECTED WORKMANSHIP) FOR UNREINFORCED CONCRETE MASONRY DESIGN TYPE M MORTAR fm = 1350 psi; Mo = 2500 psi TYPE NORMAL AND OBE LOAD COMBINATIONS SSE LOAD COMBINATIONS OF MASONRY CRITERIA USED (a) SEB INTERIM (b) CRITERIA USED (a) SEB INTERIM (b) NUMBER STRESS UNIT* ON THE PROJECT CRITERIA REV. 1 ON THE PROJECT CRITERIA REV. 1 1. Tension Perpendicular to H 23 25 38 32 Bed Joints 1tF S&G 39 40 65 52 2. Tension Parallel to Bed H 46 50 77 75 Joints 11tF S&G 78 75 130 112 3. Shear H 34 40 57 52 S&G 34 40 57 52 4. Flexure Compressive Stress All 405 445 676 1112 Fm 5. Bearing: On Full Area All 337 337 563 842 On 1/3 rd area or less All 506 506 845 1265
- 6. Reinforcement 0.5 Fy 0.5 Fy 0.9 Fy 0.9 Fy Jt. Wire Fy = 65 ksi 30,000 psi 30,000 psi
_________________________
(a) Criteria used on the project is compatible with NCMA-1974.
- H = Hollow Concrete Masonry (b) SEB Interim Criteria Rev. 1 is compatible with ACI 531-79. S = Solid Concrete Masonry G = Grouted Concrete Masonry
B/B-UFSAR 3.8-103 TABLE 3.8-16 FLEXURAL STRENGTH HORIZONTAL SPAN, NONREINFORCED CONCRETE MASONRY WALLS MODULUS OF SAFETY RUPTURE FACTOR MORTAR LOADING NET AREA ACTUAL/ CONSTRUCTION TYPE TYPE psf (psi) ALLOWABLE 8 inch Monowythe N Uniform 127 102 4.13 Hollow, 3-Core N Uniform 136 141 4.41 (Reference 1) N Uniform 127 132 4.13 N Uniform 169 176 5.50 N Uniform 173 180 5.63 O Uniform 123 128 4.00 O Uniform 158 164 5.13 8-inch Monowythe N Uniform 149 155 4.84 Hollow Joint N Uniform 160 166 5.19 Reinforced @ 16 inch N Uniform 193 201 6.28 center to center O Uniform 150 156 4.88 (Reference 1) O Uniform 186 193 6.03 8-inch Monowythe N Uniform 203 211 6.59 Hollow Joint N Uniform 196 204 6.38 Reinforced @ 8 inch O Uniform 202 210 6.56 center to center O Uniform 195 203 6.34 (Reference 1) 8-inch Monowythe N 1/4 pt 56 58 1.81 Hollow (Reference 2) N 1/4 pt 38 39 1.22 N 1/4 pt 61 63 1.97 N 1/4 pt 60 62 1.94 N 1/4 pt 69 71 2.22 N 1/4 pt 93 96 3.00 8-inch Monowythe M Center 199 217 4.72 Hollow, 2-Core M Center 176 192 4.17 (Reference 3) M Center 151 165 3.59 4-2-4 Cavity Wall, M Center 111 210 4.57 Hollow Units M Center 135 255 5.54 (Reference 3) M Center 95 180 3.91 8-inch Monowythe M Center 159 173 3.76 Hollow 2-Core M Center 159 173 3.76 Joint Reinforced M Center 191 208 4.52 @ 8" center to center (Reference 3)
B/B-UFSAR 3.8-104 REVISION 1 - DECEMBER 1989 TABLE 3.8-16 (Cont'd) MODULUS OF SAFETY RUPTURE FACTOR MORTAR LOADING NET AREA ACTUAL/ CONSTRUCTION TYPE TYPE psf (psi) ALLOWABLE 4-2-4 Cavity of M Center 159 300 6.52 Hollow Units Tied M Center 159 300 6.52 with Joint M Center 159 300 6.52 Reinforced @ 8" center to center (Reference 3) 4-inch Hollow N Center 138 365 11.41 Monowythe N Center 157 415 12.97 (Reference 4) N Center 101 268 8.38 8-inch Hollow M Center 268 202 4.39 Monowythe M Center 314 237 5.15 (Reference 4) M Center 314 237 5.15 8-inch Hollow N Center 277 210 6.56 Monowythe N Center 314 237 7.41 (Reference 4) N Center 314 237 7.41 8-inch Hollow O Center 259 195 6.09 Monowythe O Center 277 210 6.56 (Reference 4) O Center 277 210 6.56 8-inch Hollow M Center 268 202 4.39 Monowythe M Center 297 224 4.87 (Reference 4) M Center 277 210 4.56 8-inch Hollow N Center 277 210 6.56 Monowythe N Center 259 195 6.09 (Reference 4) N Center 297 224 7.00 8-inch Hollow O Center 360 271 8.45 Monowythe O Center 297 224 7.00 (Reference 4) O Center 268 202 6.31 12-inch Hollow N Center 352 142 4.44 Monowythe N Center 314 127 3.97 (Reference 4) N Center 333 134 4.19 _________________________
a) Total number of tests without joint reinforcement = 43 Average safety factor (SF) = 5.3 b) Total number of tests with joint reinforcement = 15 Average safety factor (SF) = 5.6 B/B-UFSAR 3.8-105 TABLE 3.8-16 (Cont'd)
REFERENCES:
- 1. Hedstrom, R. O., "Load Tests of Patterned Concrete Masonry Walls," Proceedings, American Concrete Institute, Vol. 57, p. 1265, 1961. 2. Fishburn, Cyrus C., "Effect of Mortar Properties on Strength of Masonry Monograph 36," National Bureau of Standards, 1961. 3. Cox, F. W., and Ennenga, J. L., Transverse Strength of Concrete Block Walls, "Proceedings," ACI, Vol. 54, p. 951, 1958. 4. Livingston, A. R., Mangotich, E., and Dikkers, R. "Flexural Strength of Hollow Unit Concrete Masonry Walls in the Horizontal Span," Technical Report No. 62, NCMA, 1958.
B/B-UFSAR 3.8-106 TABLE 3.8-17 SUMMARY OF MASONRY WALLS WITH STEEL COLUMNS BASED ON DESIGN PARAMETERS FOR BYRON UNIT 1 TYPE APPROXIMATE NUMBER OF MASONRY WALLS WITH STEEL COLUMNS OF GROUP 1 GROUP 2 GROUP 3 GROUP 4 ALL ASPECT A*>75 A*=38-75 A*=25-37 A*<25 VALUES RATIO h/s I*>500 I*=100-500 I*=50-99 I*<50 OF A* AND I* 1 2.0 47 19 10 12 88
>2.0 2 and 6 8 2 3 19 3.0 3 >3.0 3 - - 4 7 All Types All values 56 27 12 19 114
B/B-UFSAR 3.8-107 TABLE 3.8-18 AMPLIFICATION FACTORS FOR TYPICAL EXPANSION ANCHOR BASE PLATES WITH WEDGE TYPE ANCHORS MAX. ANCHOR MAX. ANCHOR REACTION REACTION PLATE ANCHOR (FLEXIBLE PLATE (RIGID PLATEAMPLIFICATIONNO. ASSEMBLY LOAD ANALYSIS) ANALYSIS) FACTOR 1 9 x 9 x 1/2 in. 12.8 k (tension) 3.2 3.2 1.0 4 anchors 1/2 in. 43.6 in-k (moment)2.85 3.2 1.0 2 9 x 15 x 1/2 in. 33.7 k (tension) 7.0 5.62 1.25 6 anchors 1/2 in. 199 in-k (moment) 7.0 6.62 1.06 3 21 x 21 x 7/8 in. 114 k (tension) 16.0 14.25 1.12 8 anchors 3/4 in. 900 in-k (moment) 16.0 14.7 1.09
B/B-UFSAR 3.10-1 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.10.1 Seismic Qualification Criteria 3.10.1.1 NSSS Equipment Seismic design criteria for engineered safety features electrical equipment require that this equipment perform its safety function for both the operating-basis earthquake and the safe shutdown earthquake. Likewise, environmental criteria presented in Section 3.11 require engineered safety features equipment to perform its safety function in the environment which would be characteristic of postulated accidents. The implementation of these criteria provides a high degree of assurance that the equipment will perform its required function in different sequences of conditions. The following list identifies the instrumentation and electrical equipment which require seismic qualification: a. pressure transmitters and differential pressure transmitters, b. process control equipment cabinets,
- c. solid-state protection system cabinets, d. nuclear instrumentation system cabinets,
- e. safeguards test racks,
- f. resistance temperature detectors, g. instrument supply inverters,
- h. reactor trip switchgear, i. power range neutron detectors, j. incore thermocouple system,
- k. main control board, l. Class 1E equipment (BOP),
- m. supporting structures - panels (BOP),
- n. electrical equipment supports (BOP), and
- o. cable tray supports (BOP). The seismic qualification testing program which will be implemented for NSSS equipment supplied by Westinghouse is specified in Reference 1. According to Regulatory Guide 1.89 (Reference
B/B-UFSAR 3.10-2 10), equipment for plants in the stage of construction permit application and having an issue date for the Safety Evaluation Report after July 1, 1974, such as the Byron/Braidwood Stations, will take into account aging effects prior to seismic qualification as specified in IEEE 323-1974. Subsection 3.11.2 presents the commitment to meet IEEE 323-1974. The seismic tests conform to the procedures specified in IEEE 344-1975 which account for multiaxis and multifrequency effects of seismic excitation as well as fatigue effects caused by a number of OBE events. This commitment was satisfied by implementation of the final Staff-approved version of Reference 1. Westinghouse has previously type tested and qualified items "a" through "h" including instrumentation included in the cabinets, to IEEE 344-1971. Item "i" is discussed in Subsection 3.10.2. Reference 2 presents the Westinghouse testing procedures used to qualify equipment by type testing. Seismic qualification testing of this equipment to IEEE 344-1971 is documented in References 3 through 9.
The main control board (item k), including the main control room control panels and remote shutdown panels, are qualified by a combination of testing and analysis. This qualification satisfies the practices recommended by IEEE 344-1975 and is documented in References 11 and 12. Items "1" through "o" are not supplied by Westinghouse. 3.10.1.2 Balance-of-Plant Equipment For purposes of the following discussion, Seismic Category I electrical power equipment is synonymous with Class 1E equipment as identified in IEEE 308-1971. The loads acting on Class 1E equipment are listed in Table 8.3-1. Class 1E power components are also identified in this table.
For purposes of the following discussion, Seismic Category I electrical and electromechanical instrumentation is synonymous with Class 1E instrumentation as defined in IEEE 308-1974. This instrumentation is listed in Subsection 7.1.1.
The seismic design criteria for Seismic Category I electrical equipment and instrumentation are as follows: a. They are designed to withstand, without the loss of nuclear safety function, safe shutdown earthquake forces and any other applicable loads transferred to the floor on which they are located. b. Equipment possessing stationary (passive) safety functions (e.g., cable supports, instrument supports, and other components which do not perform a mechanical motion as part of their safety function) is designed to ensure the operability of safety-related equipment.
B/B-UFSAR 3.10-3 c. Equipment possessing nonstationary (active) safety functions (e.g., switches, motor-operators, and other equipment which perform a mechanical motion as part of their safety function) is designed such that the operability of the equipment is demonstrated during and after the simulated seismic event by analysis, testing, or a combination of both. 3.10.1.2.1 Cable Tray and Bus Duct Supports Criteria The following criteria are used in the design of cable trays and bus duct supports:
- a. Regardless of cable tray or bus duct function, all supports are designed to meet the requirements of Seismic Category I structures by dynamic analysis using the appropriate seismic response spectra. b. The analytical maximum values are obtained by taking the square root of the sum of the squares of the stresses and reactions of all significant modes. c. Cable tray loading of 45 lb/ft2 is used throughout the design regardless of tray height or weight. 3.10.2 Seismic Analysis Testing Procedure and Restraint Measures 3.10.2.1 NSSS Compliance Seismic Category I Westinghouse-supplied instrumentation and electrical equipment were seismically tested using sine beat inputs to each of three perpendicular axes independently applied according to the procedures of IEEE 344-1971, Section 3.2. At the time of this testing, which is reported in References 3 through 9, implementation of the IEEE 344-1971 testing method fulfilled all seismic qualification requirements. The results show that there were no electrical irregularities that would leave the plant in an unsafe condition.
In the reported tests, the equipment operated properly during and after testing with equivalent ground accelerations at zero period ranging up to 0.4g and higher. The sine beat inputs were applied not only at the equipment natural frequencies but also at many frequencies (spaced at about 1/2 octave) below 33 hertz to ensure that the equipment would function normally regardless of uncertainties of building or equipment natural frequencies. The sine beat test is severe because it excites the resonant response of the equipment thereby producing the most damaging effect to the components. This test not only excites the component to motion greater than the input but also produces fatigue damage well above that produced by seismic B/B-UFSAR 3.10-4 REVISION 7 - DECEMBER 1998 disturbances. This method assumes that building natural frequency coincides with that of the equipment and is as conservative as the one proposed by the NRC staff. Any possible coupling effect loses importance when compared to the excitation of components at sensitive frequencies as is done by the sine beat test. This test therefore provides more positive proof of equipment capability than the simultaneous random input test which, because of phase relationships, could result in less severe application of the seismic input. The nuclear instrumentation system power range neutron detector has been tested in both the horizontal and vertical directions. Current, resistance, and capacitance checks were made before and after the tests. No significant changes were observed and no mechanical damage was noted. Additional tests of the power range neutron detectors were conducted using multi-frequency, multiaxial excitation as specified in Reference 1.
Equipment for a particular plant is procured on a similar basis to that which is qualified. Any equipment design changes are evaluated to determine if the changes were of a nature that could affect the results of the seismic tests. If it is determined that the changes may affect the seismic characteristics of the equipment, then the equipment is requalified for seismic integrity.
The criteria and verification procedure employed to account for the possible amplified design loads (frequency and amplitude) for Westinghouse-supplied safety-related instrumentation and electrical equipment is presented in the references of Section 3.10 (specifically Reference 2 and Appendix B of Reference 3). 3.10.2.2 Balance of Plant Compliance With IEEE 344-1971 and 344-1975 The balance of plant Class 1E equipment meets the requirement that the seismic qualification should demonstrate the capability to perform the required function during and after the safe shutdown earthquake (SSE). Both analysis and testing were used, but most equipment was qualified by testing. Analysis was used to determine the adequacy of mechanical strength (mounting bolts, etc.) after operating electrical capability was established by testing.
- a. Analysis The balance of plant Class 1E equipment with primary mechanical safety functions (pressure boundary devices, etc.) was analyzed since the passive nature of its critical safety role usually made testing impractical. Analytical methods sanctioned by IEEE 344-1975 were utilized in such cases. Refer to Table 3.10-1 for indication of which items were qualified by analysis (for complete listing of components, see the B/B-UFSAR 3.10-4a REVISION 9 - DECEMBER 2002 equipment/components identified as safety related in the Engineering controlled equipment/component database(s).
B/B-UFSAR 3.10-5 b. Testing The balance of plant Class 1E equipment having a primary active electrical safety function was tested in compliance with IEEE 344-1971 or -1975, depending on the time the test was performed. The majority of the equipment has been tested as per the 1975 issue of IEEE-344. 3.10.3 Methods and Procedures of Analysis or Testing of Supports of Electrical Equipment and Instrumentation 3.10.3.1 NSSS Equipment The adequacy of Westinghouse-supplied electrical equipment supports is verified by testing conducted according to the procedures outlined in Subsection 3.10.2. 3.10.3.2 Balance of Plant Equipment 3.10.3.2.1 Design of Cable Trays a. Seismic Qualification Criteria 1. Applicable codes, standards, and specifications a) AISI "Specification for Design of Cold-Formed Steel Structural Members," 1968 Edition. 2. Loads a) dead load D: 45 psf including the weights of cables and cable trays; b) live load L: 200 pounds at any location between the two supports, during construction; c) severe environmental load E: peak acceleration of 4% critical damping floor spectra generated by the operating-basis earthquake; and d) extreme environmental load E': peak acceleration of 7% critical damping floor spectra generated by safe shutdown earthquake.
B/B-UFSAR 3.10-6 REVISION 1 - DECEMBER 1989 3. Load combinations a) D + L, b) D + E, and c) D + E'. b. Method of Analysis and Design The equivalent static load corresponding to the peak acceleration of the floor spectra is used for analysis and design.
- c. Procedures of Analysis and Design 1. Calculate the allowable span of the cable tray subjected to dead load and live load; use 1.0 times the allowable stresses. 2. Calculate the allowable span of the cable tray subjected to dead load, vertical seismic load, and horizontal seismic load; use 1.6 times the allowable stresses, with minimum factor of safety equal to 1.05. 3. The stresses due to vertical and horizontal seismic load are combined by using SRSS (square root of the sum of the squares) method. 4. The stresses due to dead load and seismic load are combined linearly. 5. Computer program utilized: SEISHANG - Seismic Analysis of Hangers (see Appendix D). 3.10.3.2.2 Design of Cable Tray, Nonsegregated Bus Duct, and Conduit Supports a. Seismic Qualification Criteria 1. Applicable codes, standards, and specifications a) AISI "Specification for Design of Cold-Formed Steel Structural Members," and b) AISC "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," 1969 Edition. c) AWS Dl.l "Structural Welding Code." Clarifications to, and deviations from portions of AWS Dl.l are made based on engineering B/B-UFSAR 3.10-7 evaluations. Visual weld inspection requirements are based on guidelines in a document prepared by the Nuclear Construction Issues Group, NCIG-01, Revision 2, "Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants." 2. Loads a) Dead load D: loads including the weights of the cables, cable trays, conduits, nonsegregated bus ducts, and supports.
b) Severe environmental load E: operating-basis earthquake response spectra of 4% critical damping at the floor where the support is erected.
c) Extreme environmental load E': safe shutdown earthquake response spectra of 7% critical damping at the floor where the support is erected. 3. Load combinations a) D b) D + E c) D + E' b. Method of Analysis 1. Response Spectrum Method c. Procedures of Analysis and Design 1. The masses are lumped at nodes. 2. For cable tray and bus duct supports, the following applies: a) For standard supports, three structural modes, i.e., one from each direction, are used for dynamic analysis.
b) For special supports, all the structural modes up to a frequency of 33 hertz are considered.
B/B-UFSAR 3.10-8 3. For conduit supports, the following applies: a) For trapeze type supports, item 2 applies.
b) For other types of supports such as cantilevers, direct-mounted types, etc., the peak acceleration of the floor spectra is used in the manual analysis. 4. Seismic excitation from vertical and two horizontal directions is considered. The stresses due to seismic loads from different directions are combined by the SRSS (square root of the sum of the squares) method. 5. The stresses due to dead loads and seismic loads are combined linearly.
- 6. The allowable stresses for dead load only are 1.0 times the allowable. The allowable stress for dead load plus seismic load is 1.6 times the allowable stresses, with minimum factor of safety equal to 1.05. 7. The allowable slenderness ratios for compression members are given in Appendix D, Table D-45.
- 8. Computer programs utilized: a) SEISHANG, Seismic Analysis of Hangers (see Appendix D) for supports, and b) PIPSYS, Integrated Piping Analysis System (see Appendix D) for supports. 3.10.3.3 NSSS - Testing/Analysis The adequacy of Westinghouse-supplied electrical equipment supports is verified by testing conducted according to the procedures outlined in Subsection 3.10.2. 3.10.4 Operating License Review The results of the tests and analyses to demonstrate adequate seismic qualification and implementation of proper criteria are presented in Subsection 3.10.2.
B/B-UFSAR 3.10-9 REVISION 5 - DECEMBER 1994 3.10.5 References 1."Environmental Qualification of Westinghouse Class 1EEquipment," WCAP-8587, Rev. 1, September 1977. 2.A. Morrone, "Seismic Vibration Testing with Sine Beats,"WCAP-7558, October 1971. 3.E. L. Vogeding, "Seismic Testing of Electrical and ControlEquipment," WCAP-7817 (Non-Proprietary), December 1971. 4.E. L. Vogeding, "Seismic Testing of Electrical and ControlEquipment (WCID Process Control of Equipment)" WCAP-7817, Supplement 1, December 1971. 5.L. M. Potochnik, "Seismic Testing of Electrical and ControlEquipment (Low Seismic Plants)," WCAP-7817, Supplement 2, December 1971. 6.E. L. Vogeding, "Seismic Testing of Electrical and ControlEquipment (Westinghouse Solid State Protection System) (Low Seismic Plants)," WCAP-7817, Supplement 3, December 1971. 7.J. B. Reid, "Seismic Testing of Electrical and ControlEquipment (WCID NUCANA 7300 Series) (Low Seismic Plants)," WCAP-7817, Supplement 4, November 1972. 8.E. L. Vogeding, "Seismic Testing of Electrical and ControlEquipment (Instrument Bus Distribution Panel)," WCAP-7817, Supplement 5, March 1974.
9.E. K. Figenbaum and E. L. Vogeding, "Seismic Testing ofElectrical and Control Equipment (Type DB Reactor Trip Switchgear)," WCAP-7817, Supplement 6, August 1974. 10.U.S. Nuclear Regulatory Commission, "Qualification of Class1E Equipment for Nuclear Power Plants," Regulatory Guide 1.89. 11.B. F. Maurer, "Seismic Qualification of the Byron/BraidwoodMain Control Board," WCAP-10393 (Proprietary Class 2), August 1983.
12.B. F. Maurer, "Seismic Qualification of the Byron/BraidwoodMain Control Room Control Panel and Remote Shutdown Panels," WCAP-10412 (Proprietary Class 2), October 1983.
B/B-UFSAR 3.10-10 REVISION 9 - DECEMBER 2002 TABLE 3.10-1 STATUS OF QUALIFICATION (IEEE-344) DOCUMENTS FOR ELECTRICAL COMPONENTS INCLUDESCLASS1ESEISMIC SEISMICELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-11 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMIC SEISMICELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-12 REVISION 8 - DECEMBER 2000 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMIC SEISMICELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-13 REVISION 6 - DECEMBER 1996 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMIC SEISMICELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-14 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMIC SEISMICELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-15TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-16TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-17TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-18TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-19TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-20 REVISION 1 - DECEMBER 1989 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-21TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-22TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-23 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-24 REVISION 1 - DECEMBER 1989 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-25 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-26 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-27 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-28 REVISION 1 - DECEMBER 1989 TABLE 3.10-1 (Cont'd) INCLUDESCLASS1ESEISMICELEC.EQUIP. QUALIFICA- SPECIFICATIONEQUIPMENT(8)(YES/NO)VENDORLOCATIONTION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-29 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS 1E SEISMIC SEISMIC ELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-30 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS 1E SEISMIC SEISMIC ELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-31 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS 1E SEISMIC SEISMIC ELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-32 REVISION 2 - DECEMBER 1990 TABLE 3.10-1 (Cont'd) INCLUDESCLASS 1E SEISMIC SEISMIC ELEC.EQUIP. QUALIFICA- QUALIFICA- SPECIFICATIONEQUIPMENT(8) (YES/NO) VENDOR LOCATION TION LEVEL TION METHOD STATUS(4)(5)
B/B-UFSAR 3.10-33 REVISION 9 - DECEMBER 2002 NOTES 1.Where "varies" is stated for location, equipment is locatedin several areas in plant. In these cases, maximum levelsare given to envelope all such areas.2.Mounted in piping - not mounted rigidly to structure;qualification levels are based on the maximum response ofthe pipe or duct.3.The equipment under this group is non-safety-related. (Seethe equipment/component classifications in the Engineeringcontrolled equipment/component database(s).4.The seismic qualification can be verified throughcalculations that are related to equipment/components in theEngineering controlled equipment/component database(s).5.Completed as of May 16, 1990.6.Equipment is Seismic Category 1 but it does not includeClass 1E electrical equipment power. The seismicqualification can be verified through calculations that arerelated to equipment/components in the Engineeringcontrolled equipment/component database(s).7.Power and control cables are not required to be seismicallytested for IEEE Standard - 383.8.Electrical portion only.9.The cubicle cooler units contain both non-Class 1E coolingcoils and also Class 1E fan-motor assemblies. The cubiclecooler units were seismically qualified, excludingfan-motor assemblies. The fan-motor assemblies wereseismically qualified independently and are listedseparately in Table 3.10-1.