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{{#Wiki_filter:__ ~* A p y1 rw b gu { was SU"2 ng'p'[ LIMERICK GENERATING STATION t UNITS 1 & 2 i W # s N.t PSAR PAGE REVISIONS The following PSAR pages, tables and. figures are to be inserted in your copy of the PSAR. These parles are either replacements or new pages as indicated belcw: I Delete Substitute Volume 2 l Page 3.6-4 Page 3.6-4 dated July 30, 1971 Page 3.7-1 Page 3.7-1 dated July 30, 1971 Page 5.2-6 Page 5.2-6 dated July 30, 1971 Pig. 5. 2. 5 Fig.
- 5. 2. 5 dated July 30, 1971 Page 6-1 Page '5-1 dated July 30, 1971 Page 6.5-24 Page ti.5-24 dated July 30, 1971 Page 6.5-26.
Page ti.5-26 dated July 30, 1971 g (New) Page (i.7-1 dated July 30, 1971 (New) Page (i.7-2 dated July 30, 1971 volume 3 M Page 7-1 Page 7-i dated July 30, 1971 Page 7.2-23a Page 7.2-23a dated July 30, 1971 (New) Page 7.2-23b dated July 30, 1971 Page 7.3-8 Page 7.3-8 dated July 30, 1971 Page 7.3-15b Page 7.3-15a dated July 30, 1971 Page 7.3-16a Fig. 7.7.lc Fig. 7.7.le dated July 30, 1971 Fig. 7.7. le Fig. 7.7. le dated July 30, 1971 Volume 4 Fig. 12. 2. 2 Fig. 12.2.2 dated July 30, 1971 Fig. 12.2.3 Fig. 12.2.3 dated July 30, 1971 Fig. 12. 2. 4 Fig. 12.2.4 dated July 30, 1971 34G8 8709280273 870921 PDR FOIA PDR MENZ87-111
PSAR PAGE REVISIONS (Continued) .\\ j Delete Substitute Volume 5 Page A.3-3 Page A.3.3 dated July 30, 1971 (New) Page A.3. 3a dated July 30, 1971 Page C-i Page C-1 Page C-il Page C-li Page C.4-1 Page C.4-1 dated July 30, 1971 through through C.4-9 C.4-23 Supplement 3 Page 10 Page 10 dated July 30, 1971 Page 10a Page 10a dated July 30, 1971 i \\ 0-
] I,Gs the fuel depletion process with spatial neutron flux and energy distributions typical of reactor operating conditions. At selected burnup intervals, the nuclide concentrations are used to recalculate revised flux and material weighted cross sections with the lattice model and these are again recycled through two dimensional diffusion theory. A large capacity, nodal, three dimensional boiling water reactor simulation code which provides for representation and calcu-lation of spatially varying voids, control rods, burnable poi-sons and other variables is used to compute three-dimammional power distributions, exposure and reactor thermal-hydraulic characteristics at the beginning of core life and as burnup progresses. In addition, it can serve to determine control red strategy through life, power response to changes in core flow and to calculate assembly as we41 as core exposure. 3.6.5.2 Reactivity Control The excess reactivity designed into the initial core is con-trolled with a control rod system supplemented by gadolinia-urania burnable poison rods. The core is designed to permit the energy extraction of 19,000 MWD /T of uranium averaged over the initial core load. This exposure can be achieved with the reactor operhting at full power at the end of each cycle. The average fuel enrichment for the initial core load is chosen to provide excess reactivity in the fuel assemblies sufficient to overcome the neutron losses due to core neutron leakage, moderator heating and boiling, fuel temperature rise, and equilibrium xenon and samariusa poison-ing; also included is an allowance for fuel depletion. Fol-lowing the initial cycle, more new fuel may be added to achieve annual refueling during the desired refueling month. During fuel burnep, control rods are used, in part to counter-act the power distribution effect of steam voids indicated by the in-core flux monitors. In combination, the control rod and void distributions can be used to flatten gross power beyond that which is possible in a non-boiling core. The design provides considerable flexibility in the control of gross power distribution. This permits regulation of fuel burnup and isotopic composition throughout the core to the l extent necessary to counteract the effect of voids on axial power distribution at the end of a fuel cycle, when few control rods remain in the core. The control rod cystem is designed to provide adequate control of the maximum excess reactivity anticipated during the equilibrium fuel cycle operation. The initial core loading, however, has an excess reactivity somewhat higher than that of the equilibrium core. Thus, the design basis for the initial burnable poison loading is that it shall compensate the reactivity difference between the control rod system 3.6-4 July 30,1971 e
' LGS 3.7 THERMAL AND HYDRAULIC ')ESIGN 3.7.1 Power Generation Objective The~ objective of the thermal and hydraulic design of the core is to achieve power operation of the fuel over the life of the core without sustaining fuel damage. 3.7.2 Power __ eration Design Bases 1. The thermal hydraulic characteristics of the core shall provide the ability to achieve rated core power output throughout the design lifetime of the fuel without sus-taining fuel damage. 2. The thermal hydraulic characteristics of the core shall provide thermal margin which, in conjunction with the plant equipunant characteristics, manclear instrumentation, and the reactor protection system assures that no fuel damage will occur during normal operation or abnormal operational transients caused by reasonably expected single operator error or equipment malfunction. 3.7.3 Safety Design Bases 1. The thermal hydraulic design of the core shall establish limits for use' in setting devices of the Nuclear Safety Systems so that no fuel damage occurs as a result of ab-normal operational transients. (See Section 14," Plant Safety Analysis".) 2. The thermal hydraulic design of the core shall establish a thermal hydraulic safety limit for use in evaluating the safety margin relating to the public safety conse-quences of fuel barrier failure. 3.7.4 Thermal and Hydraulic Limits 3.7.4.1 Steady state Limits For purposes of maintaining adequate margin during normal steady state operation, the minimum critical heat flux ratio (HCHFR) is maintained in excess of 1.9 relative to the design correlation limit lines (ref.1); the maximum linear heat gener-atson rate is maintained below 18.5 kilowatts per foot. Oper-ating power and peaking factors are not specified; these parameters are determined subject to a number of constraints, including the thermal limits noted previously. To accommodate uncertainties and to assure that no fuel damage results even during the wors t anticipated transient condition at any time i in life, the core and fuel design basis for steady state operation (i.e., MCHFR 1.9 and LHGR 18. 5 kw/f t) has been i 3.7-1 July 30, 1971
/ LGS f. TIP System guide tubes are provided with an isolation valve which closes automatically upon receipt of proper signals and after the TIP cable and fission chamber have been retracted. In series with this isolation valve, an additional or backup isolation shear valve is included to assure integrity of the containment in the unlikely event that either the isolation valve should fail to close or the chamber drive cable should fail to retract if it should be extended in the guide tube during the time that con-tainment iso 1'ation is required. g. Isolation valves are either actuated by 'various sig-nals or are remote manually operated, as appropriate. Refer to Table 7.3.1, Lines Penetrating the Primary C ntainment. 5.2.3.5 Primary Containment Venting and Vacuum Relief System The primary containment is vented as required to eliminate pressure fluctuations caused by temperature changes during various operating modes. This is accomplished through venti-lation purge connections which are normally closed while the 0 reactor is at a temperature greater than 212 F. The suppres-I sion chamber is vented separately. (Refer to Figure 5.3.1, Reactor Building Ventilation and Standby Gas Treatment.) 5.2.3.6 Primary Containment Cooling and Ventilation System 1 The Primary Containment (Drywell) Cooling System utilizes fan- 't coil units distributed inside the drywell. (Refer to Figure 5.2.8, Drywell Cooling.) Each fancoil unit consists of two cooling coils and two direct connected, two speed motor-driven fans. Each cooling coil is connected to a separate water sup ply and return piping system inside the drywell to permit the use of either the Chilled Water System for normal service or the Reactor Building Cooling Water System for standby service. Should the Chilled Water System malfunction, the standby sys-tem continues to cool the drywell. Each fancoil unit is manu-ally controlled from outside the primary containment. Each of the two f ans in a fancoil unit may run individually or simul-taneously. Drywell space temperatures, and inlet and outlet temperatures of the f ancoil units, are indicated outside the primary containment. High fan discharge terporature is annun-ciated in the main control room. Each f ancoil units are designed to run at 340 F and 55 psig l high radiation environment and will operate under ony one of the following modes: 5.2-6 July 30, 1971
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LGS SECTION 6 CORE STANDBY COOLING SYSTEMS CONTENTS 6.1 SAFETY OBJECTIVES 6.2 SAFETY DESIGN BASES 6.3 stb 4 MARY DESCRIPTION
6.4 DESCRIPTION
6.4.1 High Pressure Coolant Injection System 6.4.2 Automatic Depressurization System 6.4.3 Core Spray System 6.4.4 Low Pressure Coolant Injection system 6.5 S FETY EVALUATION 6.5.1 Sussary 6.5.2 Performance Analysis 6.5.2.1 Analysis Model 6.5.2.2 High Fressure Coolant Injection System j 6.5.2.3 Auto Depresturization System 6.5.2.4 Core Spray System 6.5.2.5 Low Pressure Coolant Injection System 6.5.3 Integrated Operation of the Core Standby Cooling Systems 6.5.3.1 Liquid Line Breaks 6.5.3,2 Steam Line Breaks 6.5.3.3 Effect of Fuel Clad Failure on Core Cooling 6.5.4 Core Standby Cooling Systems Redundancy 6.6 INSPECTION AND TESTING 6.7 CORE STANDBY COOLING SYSTEM IMPROVEMENTS 6.7.1 HPCI System Improvements 6.7.2 LPCI System Improvements 6-1 July 30, 1971 )
LGS' taken for the pressure reduction effect of the cold LPCI water in the reactor vessel. The effective range of LPCI alone (3 or 4 pumps) for the spectrum of steam or liquid line breaks is shown in Figure 6.2.1. The half-widt'n portion of the bar shows the overlap with other Core Standby cooling Systems. To assure continuity of core cooling, signals to isolate I the primary or secondary containments do not operate any LPCI valves. This arrangement satisfies safety design basis 6. The two testable check valves (one in each loop) are the only I.PCI equipment in. the primary containment required to actuate during a loss-of-coolant accident which require consideration for the high temperature and humidity environment in the containment from the accident. The type of valve chosen actuates on flow through the pipeline, independent of any external signal. The actuator is provided only for test. Thus, neither the normal nor accident environment in containment affects the operability of the Low Pressure Coolant Injection equipment for the accident. It is concluded that safety design basis 9 is satisfied. Using the suppression pool as the source of water for LPCI establishes a closed loop for recirculation of LPCI water escaping from the break. It is concluded that safety design basis 11 is satisfied. The LPCI and appropriate portions of the recirculation loops are designed as Class I Nuclear so that they meet design basis 8. 6.5.3 Integrated operation of the Core Standby Cooling Systems s The previous discussion has described the performance and operation of each of the CSCS individually. This discussien is directed toward the integrated performance of the CSCS, i. e., how the CSCS operate together to provide core cooling for the entire spectrum of loss-of-coolant accidents, vi:., a break of a liquid line and a break of a steam line. The primary emphasis of the discussion i 6.5-24 July 30, 1971 l
LGS 6.5.3.1 Liquid Line Breaks 1. Large Breaks The double-ended recirculation line break is one of the. bases for the design of the Core. Standby Cooling Systems and the containment response calculations. The containment response is discussed in Section 14, " Plant Safety Analysis". This accident is analyzed with a nine node reactor volume simulatic,n model. The nine nodes are the lower plenum, the core, the upper plenum, the leakage region, the separation zone, the steam. dome, the downcomer and the two recirculation loops. The jet pump modeling assumes that conservation of momentum i and drive pump trip can be included. The core and leakage regions each have 10 common pressure subnodes. Included in the model is a method of calculating the movements of the liquid level in the separator region. A vapor to liquid relative velocity of I ft/sec is assumed in these calculations (refs. 3, 7). 1 It is assumed tha't the reactor is operating at design power when a complete circumferential j rupture instantly occurs in one of the two recirculation system suction lines. An inter-lock assures that the valve in the equalizer l line between the jet pump headers will be closed when both recirculation pumps are operating; thus, the area available for coolant discharge from the reactor vessel would be the sum of 10 jet pump nozzle areas and the cross sectional area of the main recirculation line and the reactor water 2 cleanup suction line (4.82 f t ). This is the worst l s case for the CSCS analysis. i Immediately af ter the break, critical flow would be ~ established at the break. The large increase in core void fraction that would be caused by the decreasing vessel pressure would be sufficient to render the core suberitical. High drywell pressure would initiate mechanical scram of the control rod system in less than one second. In i about 9 seconds the liquid inventory 'in the down-l comer and the separator region of the vessel would be depleted and the break flow would switch from liquid to vapor; this would result in a large increase in the vessel depressurization rate. 6.5-26 July 30, 1971 i
/ LGS 6.7 CORE STANDBY COOLING SYSTEM IMPROVEMENTS The core standby cooling systems originally proposed for Limerick (described in Sections 6.1 through 6.5) are adequate in that the ECCS provided in depth protection with redundant systems meeting all criteria with significant margin.
- However, Philadelphia Electric will provide additional cooling' margin and system reliability for ECCS.
The following improvement to further enhance the margins already present 'will be made to the core standby cooling systems for the Limerick Units:
- 1. The HPCI system will discharge into a core spray header rather than into a feedwater line.
- 2. The HPCI system will be analyzed in detail to determine whether the present equipment can attain effectiveness as a core spray.
- 3. The LPCI system will discharge into the vessel inside
) the core shroud through four separate penetrations rather than into the recirculation piping. 6.7.1 HPCI System Improvements i Injection of the HPCI fluid through the core spray sparger results in more efficient utilization of the flow than if the water is injected outside the core shroud. The reason for this is that by injecting the water over the core, it must pass through the core before it is lost from the primary system. Thus, water levels inside the shroud would be main-tained higher for longer times which result in better core cooling and therefore further lowering of peak clad temperatures. In addition to the above improvement, the HPCI flow rate may be increased by optimal pump impeller design. The exact mag-nitude of the gain will be established during detail system design. The purpose of this increase is to maximize the prob-ability that the HPCI can function as a core spray over that section of the break spectrum for which pressure is available. If indeed the HPCI can be claimed as a core spray, this would result in peak clad temperatures remaining below 7000F for breaks up to approximately 0. 6 f t. That is, core heat up 2 for this range of break sizes would be precluded by the early actuation of core spray. 6.7-1 July 30,19 71
[ LGS 6.7.2 LPCI System Intprovements i Revising the LPCI System such that it discharges directly through four seperate nozzles into the reactor vessel eliminates the possibility of losing the LPCI mode of the RHR system due to a single component failure (such as an injection valve). This approach is a further improvement in LPCI system reliability.. In addition, the use of smaller, 3 l-faster opening valves result in significantly faster flooding time to give even greate cooling margins. l } The LPCI water injection into the top of the core instead of j-into the lower plenum provides some core cooling before the lower plenum is filled with water. However, credit has not f been taken for this cooling in any of the preliminary analy-j sis. It is expected that these improvements to the low pressure coolant injection system will result in a peak fuel temperature reduction during the worsg case accident (DBA plus single failure) of between 200 to 280 F t -l
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s a 6.7-2 July 30,1971
SECTION 7 CONTROL AND INSTRUMENTATION CONTENTS 7.1
SUMMARY
DESCRIPTION 7.1.1 Safety Systems 7.1.2 Power Generation Systems ' 7.1.3 Safety Function 7.1.4 Plant Operational Control 7.1.5 Definitions 7.2 REACTOR PROTECTION SYSTEM 7.2.1 Safety Objective 7.2.2 Safety Design Basis 7.2.3 Description 7.2.3.1 Identification 7.2.3.2 Power Supply 7.2.3.3 Physical Arrangement 7.2.3.4 Logic 7.2.3.5 operation 7.2.3.6 Scram Functions and Bases for Trip Settings 7.2.3.7 Mode Switch 7.2.3.8 Scram Bypasses 7.2.3.9 Instrumentation l 7.2.3.10 wiring 7.2.4 Safety Evaluation 7.2.5 Inspection and Testing 3 7.2.6 Additional Information l 7.3 PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM 7.3.1 Safety Objective 7.3.2 Definitions 7.3.3 Safety Design Bases 7.3.4 Description 7.3.4.1 Identification 7.3.4.2 Power Supply 7.3.4.3 Physical Arrangement 7.3.4.4 Logic 7.3.4.5 Operation 7.3.4.6 Isolation Valve Closing Devices and Circuits 7-i July 30, 1971
LGS 4 7.2.6 ADDITIONAL INFORMATION Studies have been conducted to evaluate-the effects of a postulated failure to scram under anticipated transients. The General Electric Company report NEDO-10349, March 1971, has been submitted to AEC for review. The postdlated failure of the total scram protection function is not considered a credible event and has not been our design basis. However, provisions will be made in the Limerick design such that the function of tripping the recirculation pumps, as described in the above report, will be added. I \\ l 7.2-23a July 30, 1971 j
1 LGS SUBSECTION 7.2 REACTOR PROTECTION SYSTEM REFERENCE 1. Hentschel, M. K. et al., " Compliance of Protection Systems to Industry Criteria General Electric BWR Nuclear Steam Supply System", June 1970 (NEDO - 10139) l i 1 I I s l 7.2-23b July 30,1971 e p.
LGS 7.3.4.4 Logic The basic logic arrangement is one in which the operation of an automatic isolation valve is controlled by two trip systems. Where many isolation valves close on the same signal, two trip systems control the entire group. Where just one or two valves must close in response to a special signal, two trip systems may be formed from the instruments provided to sense the special condition. W 1ves that respond to the signals from common trip systems are identified in the detailed descriptions of isolation functions. Each trip system has a pair of logics, each logic of which receives input signals from at least one channel for each monitored variable. Thus, two channels are required for each essential monitored variable to provide independent inputs to the logic of one trip system. A total of four channels for each essential monitored variable is generally provided for the logics of both trip systems except for HPCI excess flow, which is 1:2 logic. Figures 7.3. 2 and 7.3.3 illustrate typical isolation control arrangements for motor-operated valves and for the main steamline isolation valves. The actuators associated with one logic pair provide inputs into each of the actuator logics for that trip system. Thus, either of the two logics associated with one trip system can produce a trip system trip. The logic is a one-out-of-n arrangement, where n may be two or more. To initiate valve closure the actuator logics of both trip systems must be tripped. The overall logic of the system could be termed one-out-of-two taken twice. The basic logic arrangement just described does not apply to class C isolation valves and testable check valves. Exceptions to the basic logic arrangement are made fo.r the HPCI and RCIC isolation valves as described below. 7.3.4.5 Operation During normal operation of the plant, when isolation is not required, sensor and trip contacts essential to safety are closed; channels and trip logics are normally energized. j! Whenever a channel sensor contact opens, its auxiliary q relay deenergizes causing contacts in the trip logic to l' open. The opening of a sufficient number of contacts in the L !i II i I 'i i 7.3-8 July 30, 1971 I i
7 - - - - - - - - - - - - - - ~ = l LGS The main steam line high flow trip setting was selected high enough to permit the isolation of one main steam line for test at rated power without causing an automatic isolation of the rest of the steam lines yet low enough to permit early detection of a gross steam line break. 5. Low Steam Pressure at Turbine Inlet (Table m 7.3.1, Signal P) { I Low steam pressure upstream of the turbine stop valves while the reactor is operating could indicatu a malfunction of the pressure i regulator in which the turbine control valves or turbine bypass valves open fully. This - I action could cause rapid depras:;urization of the nuclear system. From part-load operating conditions, the rate of decrease of nuclear ) system. saturation temperature could exceed the design rate of enange of vessel temperature. l e I K*
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1 LGS design work. This equipment is designed to meet a I detailed functional requirement specification. The I design is supported by field experience and test experience. 2. Special Supplemental Methods Some complex equipment (e.g., reactor internals) is normally sized by rational stress ana2ysis techniques ] and requires supplemental criteria in areas where 1 industrial codes do not apply. A.3.1.1 Piping class I Seismic piping is classified as either rigid or flexible. Rigid piping.is that which has a fundamental fre-quency in the rigid range of the spectrum curves for the building locations. This generally corresponds to frequencies greater than about 30 cps. These piping systems are I analyzed with static loads corresponding to the noceleration l in the rigid range of the spectrum curves, j The dynamic analysis of flexible class I Seismic Piping Systems for seismic loads is performed using the spectruma response method. The percentage of critical damping for all modes is 0.5 for the Operating Basis Earthquake (OBE) and 0.5 for the Design Basis Earthquake (DBE). I The vertical and horizontal floor response spectra applied I to the Piping Systems are developed as part of the seismic f analysis for the building in which the piping is located. When the seismic load is due to the Design Basis Earthquake (0.12 g horizontal plus 0.08 g vertical), the vectorial i combination of all longitudinal primary stresses does not j exceed material yield stress at temperature unless higher j allowable limits are calculated and substantiated by the methods outlined in PSAR Volume 5, Appendix C. The main steam line (MSL) from the MSL isolation valve up to and including the turbine stop valve and the turbine bypass line from the main steam linet ', the bypass valve header, including the header, and the.. associated restraints will be designed by the use of a dynamic seismic analysis to withstand the OBE and DBE loads within the limits of the ANSI. B31.7 Clan II piping code and Appendix A. The dynanic input for design of the MSL will be derived from a time history modal analysis (or an equivalent method) of the pertinent supporting structures. E
LGS 1 l J = ~ + The Class II Turbine Building, housing,the MSL's may undergo I some plastic deformation under the DBE; however, the plastic deformation will be limithd,to a ductility factor of 2 and an elastic multi-degree of freesha, system A.na(ysis will be per-formed. The MSL aupporting structures ~(these portions of the Titchine Building) will be such that the' MSL and its supports can perf orm their safety function under the class I seismic loading conditions. The stress allowable and associated de-formation for piping will b9 limited to 1.2 times the stress allowable (S ) for OBE and yield stress for DBE. h A.3.1.2 Equipment Equipment is supported or restrained to accommodate seismic loading determined in accordance with the criteria defined in Appendix C, " Structural Loading Criteria." A.3.2 Ma terial_s,s A.3.2.1 Brittle Fracture control for Ferritic steels The fracture or note.h toughness properties and the operating temperature of ferritic materials in systems which form the j react.or coolant and primary containment pressure boundaries are controlled to ensure adequate toughness when the system is pressurized to more than 20 percent of the ) i i t A 3-3a July 30, 1971 I
gg F 'APPENDIF.C STRUCTURAL DESIGN CRITERIA CONTENTS 4 C.1 CLASSIFICATION OF STRUCTURES C.1.1 General C.I.2 class I Seismic Structures C.1.3 Clama II Seismic Structures C.2 STRUCTURAL DESIGN BASIS, C.2.1 Dead and Live Loads C.2.2 Seismic Loads C.2.3 Wind Loads C.2.4 Tornado Loads C.2.5 Loading Combinations C.2.5.1 Notations C.2.5.2 class I Seismic Structures C.2.5.3 Class II Seismic Structures C.2.6 Governing Codes and Regulations i C.3 SEISMIC ANALYSIS OF CLASS I SEISMIC STRUCTURES f C.4 N,SSS EQUIPMENT LOADING DESIGN CRITERIA C.4.1 Loading Criteria { C.4.1.1 Normal Design Techniques C.4.1.2 Core Support Structures C.4.1.3 Reactor Internal Structures other than Core Support C.4.1.4 Pressure Vessels, Piping, Pumps and valves C.4.1.5 Structural Steel C.4.1.6 Other Equipment C-i July 30, 1971
~ - - ~ ~ LGS APPENDIX C STRUCTURAL DESIGN CRITERIA TABLES Table No. Title C.2.1 Damping Factors C.2.2 Wind velocities and Pressures C.4.1 General Definitions for Loading Criteria C.4.2 Minimum Safety Factors C.4.3 Ultimate Design Stress Values for Piping and Reactor Vessel Materials C.4.4 Suppleanentary Limit Criteria for Reactor 1 Internal structures C.4.5 Core Support Structures, Stress Categories and Limits of Stress Intensity for Normal and Upset Conditions C.4.6 Core Support Structures, Stress Categories and Limits of Stress Intensity for Emergency , Conditions C.4.7 Core Support Structures, Stress Categories and Limits of Stress Intensity for Faulted Conditions C-ii July 30, 1971 =
l ~ I,GS l j C.4 NSSS EQUIPMENT LOADING JESIGN CRITERIA C.4.1 Loading Criteria class I equipment is examined to assure its ability to with-stand seismic requirements. Experienced designers determine which specific portions of systems and components require further examination. 'J'he techniques used in this determination fit into two general categories. They are: Normal analytical techniques using empirical design a. methods, as defined by appropriate design codes. b. Special techniques (aumployed to supplement code cal-culations, or to cover conditions not considered by existing codes). C.4.1.1 Normal Design Techniques All class I equipment is designed in accordance with applica-ble industrial codes. The 11mmits contained in the applicable design coden will not be exceeded. Some codes utilize empiri-cal design methods for equipment which cannot be sized by conventional rational stress analysis methods, and which do not require a detailed stress analysis for primary design work. This equipment is designed to meet a detailed functional requirement specification. The design is supported by empirical field experience and test experience. Fwa=mles are valve bodies and pump cases. C.4.1.2 Core Support Structures The stress, deformation and fatigue criteria presented in Tables C.4.5, 6 and 7 are used. These criteria are supple-mented, where applicable, by the criteria of Table C.4.4 but in no case are the criteria presented in Tables C.4.5, 6 and 7 exceeded for core support structures. C.4.1.2.1 Bolting The design stress intensity limits used in the design of bolt-ing for reactor core support structures are as follows: 1. The maximum value of the primary plus secondary membrane stress intensity, including stress from preload, averaged across the area of either the shank or threads, shall be no greater than the lesser of 90% of the yield strength or 2/3 of the ultimate strength, both at temperature. C.4-1 July 30, 1971
s LGS 2. The primary plus secondary membrane stress plus
- bending stress at the periphery of the bolt shall be no greater than the lesser of 1.2 times yield strength or 8/9 of the ultimate strength both'at temperature.
3. The average value of shear stress in the threads l ) is no greater than 0.6 of yield strength at temp-erature. I 4. The average value of bearing stress under the head j of bolt is no greater than 2.7 times yield strength l at temperature. The above stated criteria are used for normal, upset, emer-gency and faulted conditions. C.4.1.3 Reactor Internal Structures Other than Core Support The stress, deformation and fatigue criteria listed in Table C.4.4 or empirical methods such as described in paragraph C.4.1.6 are used in the design of the reactor internal struc-ture. l As noted in Table C.4.1, the loading conditions, are classified into four categories, with the plant requirements specified for each. In turn, a minimum safety factor is imposed based on the plant requirements, for example, a higher safety factor or margin between normal operation and failure is required for normal conditions where equipment must continue in operation, whereas a lower safety factor (but still gretter than 1) or margin is allowed for faulted conditions where the system is not required to remain operational, but need only shutdown safely. The safety f actors used in the loading critoria analyses are shown in Table C.4.2. It is not planned to use stress limits associated with faulted conditions as shown in Table C.4.4 for the equipment and com-ponents which (1) are not part of the reactor coolant pressure boundary and (2) are covered by applicable design codes. Table C.4.3 lists supplementary criteria which are used in the design of the station. The deformation, buckling stability and fatigue limits included in Table C.4.4 are included for completeness but are not necessarily applied to all components. Where it is clear that the fatigue, excess deformation or buck-ling limit is not applicable to a particular structure or com-ponent, a formal analysis with respect to that limit will not be performed. Two limiting criteria are considered in Table C.4.4 which negate the need for specific strain limits. These are the ') deflection limits and plastic instability limits described below. l i C.4-2 July 30, 1971
1 LGS The deflection limit requires that maximum permissible defor-mation under combination loading be limited to 80% of the loss of function (LOF) deformation (calculated on a conservative basis). As a practical matter, the stresses in most of the critical components are so low that these deformation limits are not invoked. When combination loading stresses do exceed the yield stress, the plastic instability design criterion would permit a maxi-zum load equal to 80% of the plastic inst. ability load. This criterion is more conservative than the recent edition of Section III of the ASME Boiler and Pressure Vessel Code which permits 90% of the plastic instability load. Using this criterion, the strain corresponding to this load varies from about 104 (non-strain hardening materials) to about 35% (strain hardening materials) of the ultimate strain at temp-erature as determined by standard ASTM tensile tests. It has not been necessary to use this criterion in the past, however, the method does represent the upper bound of strain permitted within the criterion. Primary stresses due to fault conditions are limited for design purposes to 2 Sm under combination loading. Since Sm implies a minimum factor of safety of 3 (e.g., S 1/3s ultimate) the minimum factor of safety on load m obtained from this criterion would be 1.5. The fact that the maximum load permitted is only 90% of the maximum load permitted by ASME Section III, a code generally recognized as being quite conservative, should demonstrate the adequate margin of safety present in the criteria. C.4.1.3.1 Bolting The design stress intensity limits used in the design of l bolting for reactor internal structures, other than core support, are the limits specified in Table C.4.4 for ductile metal components. C.4.1.3.2 Fatigue Limits The fatigue limit criteria for analysis shown in Table C.4.4 cre essentially identical to the fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code except that a factor of safety of 20 on cycles below the mean fatigue data curves for the material is already contained in the Section III f atigue design curvas and a cumulative usage f actor of 1.0 is permitted whereas the criteria shown in Table C.4.4 permits the use of the mean fatigue data directly with the f actor of safety of 20 being applied to the cumulative usage (i.e., usage is limited to 0.05 in Table C.4.4 rather than 1.0). This is an equivalent procedure which permits a fatigue analysis to be performed directly for materials which may not be covered by applicable industry codes and for which there are no code fatigue curves available. F e, C.4-3 Jaly 30, 1971 [
'm.a h v{$ 4 ,e a m. _.e LGS The alternate criterion for an actual fatigue test represents a margin of safety of 3 below failure or loss of function. The experimental programs would be designed so as to insure that conservatism is present in all aspects of the test including geometry, tolerances, loading conditions, etc., so that a factor of safety of 3 should be more than ample. C.4.1.4 Pressure Vessels, Piping, P - s and Valves The criteria for emergency. and fault conditions of equipment constructed in accordance with the requirements of Quality Group (See paragraph A.2) B or C codes and standards are as follows: E E B-C = A WT F B-C = A Y Where U, E and F are the criteria listed in the applicable code fof up$et, em$rgency and fault conditions, respectively, for Group A equipment, U are the criteria listed in the applicable code for upsedonditions for Group B or C eghip-ment, and E and F are the criteria to be used for emer- . gency and fIuit conditions respectively, for Group B or C equipment. Table Q 15.22 contains a list of the Quality Group B and C equipment furnished by General Electric along with the loading conditions and stress limits for design. Where analysis is required for the faulted condition on Quality Group A pumps ASME Section III stress limits will be used. Where analysis is required for the faulted condition on Quality Group A valves B31.7 Code case 70 stress limits will be used. C.4.1.5 Structural Steel Stress and deformation criteria of st.ructural steel equipment shall be 0.9 of the yield stress for emergency design conditions and either the yield stiess or plastic deformations that do not prevent accomplishment of the equipment safety functions for fault conditions. C.4.1.6 Other Equipment For other aquipment the criteria shall be based on the criteria ) established in applicable codes and standards for similar equip-ment, by manufacturers standards (e.g., turbines), or by empirical methods based on field experience and testing. c.4-4 n '- ~
3 k LGS i . i TABLE C.4.1 I GENERAL DEFINITIONS FOR LOADINC' CRITERIA t Class I Structures and Equipment - Structures and equipment l that are essential to the safe shutdown and isolation of the i reactor or whose failure or damage could result in significant release of radioactive material. t l Class II Structures and Equipment - Structures and equipment j that are important to reactor operation but are not essential to the safe shutdown and isolation of the reactor and whose 4 failure cannot result in a significant release or radioactive l j material. 4 Class III Structures and Equipment - Structures and equipment that are not essential to the operation, safe shutdown or iso-lation of the reactor and whose failure cannot result in the-release of radioactive material. Note: Class II and III items shall not degrade higher class l
- items, i
) Normal Conditions (Expected during 40-year operation) Any condition anticipated to occur in the course of operation of the plant under planned, expected conditions (in the absence of upset, emergency, or faulted conditions). j l Upset Conditions (Likely or possible during 40-year operation) l Any deviations from normal conditions anticipated to occur of ten enough that station design should include a capability I to withstand the conditions with the station remaining opera-tional or being capable of regaining its operational status. Upset conditions include: abnormal operational trancients caused by a fault in a bystem component requiring its isolation from the system; transients due to loss of load or power, any system upset not resulting in a forced outage, and operating basis earthquake. Emergency Conditions (Low probability during 40-year operation) Any deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system. Emergency conditions have a low probability of occurence but are evaluated to provide assurance that no gross loss of structural integrity will result'as a concomitant effect of any damage developed in the system. C.4-5 July 30, 1971
_ _ _ - - - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - -. - - - _. _ _. - _ _ - _ - _ - - - _ - _ - - - - - - - - - - -, - - S gg TABLE C.4.1 ' (Continued) Faulted Conditions (Extremely low probability) Extremely low probability postulated events or combinations of conditions whose consequences are such that the integrity and operability of the nuclear system may be impaired to the-extent where considerations of public health and safety are involved. Operating Basis Earthquake (OBE) - An improbable event, which s however may be considered possible durihg the 40-year station design life, and therefore an upset condition for which the station must be capable of remaining operational, or regaining its operational status. Design Basis Earthquake (DBE) - A low probability event, and therefore an emergency condition, which however is evaluated to assure station capability for safe shutdown. J Pipe Rupture - The low probability rupture of a small Class I i system pipe, which must therefore be considered an amergency condition; or the extremely low probability of rupture of a major class I system pipe - such as the recirculation line break or main steam line break - which are used as design basis accidents for safety evaluations of station capability for protecting the public health and safety; or the extremely l low probability of pipe rupture in conjunction with a design basis earthquake. Minimum Safety Factor, SF MIN Minimum safety factors appear-ing in loading criteria used for design of high reliability Class I equipment. They are based on the operational or safe shutdown requirements placed upon the station, and the nature and severity of the loading condition. \\ i i 1 C.4-6 July 30, 1971
] m-6$+. LGS ( ) h TABLE C. 4. 2 MINIMUM SAFETY FACTORS Condition (40 vr. Probability) Governing Loading Condition min SF Upset (Likely or Possible) N and Ao 2.25 --or-- N and U Emergency N and R (Iow probability) 1.5 --or-- N and Ag 1.5 i 1 --or-- t Other 2.25 to 1.5 Fcult N and AM and R 1.125 (extremely low --or - i probability) I Other 1.5 to 1.125 i Whare: N = normal loads i U = upset loads excluding earthquake i' A " Operating basis earthquake including any associated D transients AM= design basis earthquake including any associated transients R = any pipe rupture loading includf ng any associated transients 0 C.4-7 Julv in. 10'71 i
m s LGS TABLE C.4.3 Table C.4.3 ULTIMATE DESIGN STRESS __ VALUES FOR PIPING AND PRESSURE VESSEL MATERIALS 4 ITDI REVISED GE CRITERIA t Above t of Limit Normal Ultimate Emer - Ferr - PV 2.25S, 50 -50 Emer - Aust - PV 2.25Sm 50 -50 Emer - Ferr - Pipe 2.25S, 50 -50 Emer - Aust - Pipe 2.25Sm 50 -50 Pault - Ferr 1.33LLB (1.5Sm) 100 (-) 67 (-) Fault - Aust 1.33LLB (1.5Sm) 100 67
- Fault - PV 0.8 LPI 140 80
- Fault - Pipe 0.8 LPI 140 80
- Experiment 0.89 (ULT) 167 89
- These conditions will not be used prior to further discussion with the AEC Staff.
NOTE: Sm, LLB, and LPI are defined in Table C.4.4. C.4-8 July 30, 1971
LGS TABLE C.4.4 SUPPLEMENTARY LIMIT CRITERIA FOR REACTOR INTERNAL STRUCTURES PRIMARY STRESS LIMIT
- No.
Any One of (Oniv One Reauired) General Limit P1 Elastic Evaluated Primary Stresses SE < 2.25 Permissible Primary Stresses IF sr g P2 Elastic Evaluated Primary Stress SE < 0.75 Conventional Ultimate Strength EiT ~ SF,g, at Temperature P3 Elastic-Plastic Evaluated Nominal Primary Stress SEP < 0.9 Conventional Ultimate Strength at ED~ ~ SF,g, Temperature P4 Permissible Load LP < 1,5 Largest Lower Bound Limit Load EM ~ SF,g, P5* Permissible Load LP < 0.9 Plastic Instability Load EEY ~ SF,g, P6* Permissible Load LP < 0.9 Ultimate Load from Fracture EUF sr Analysis P7* Permissible Load LP or LP < l.0 Ultimate Load or Loss of Function LU
- LLF, SF
~ g min S = The tabulated value of ASME III, or its equivalent, allowable stress at temperature. Primary stresses evaluated on an elastic basis. The SE = effective membrane stresses are to be averaged through the load carrying section of interest. The simplest average bending, shear, or torsion stress distribution which will support the external loading will be added to the membrane stresses at the section of interest. SP = Permissible primary stress levels under normal or upset conditions under applicable industry code. SU Conventional ultimate strength at temperature or loading = which would cause a system malfunction as delineated in the design specification, whichever is more limiting. C.4-9 July 10. 1971
- J - -. 1 LGS TABLE C.4.4 - (Continued) PRIMARY STRESS LIMIT (Continued) SEP Elastic plastic evaluated nominal' primary stress. Strain = hardening of the material snay be used for the actual monotonic stress strain curve at the temperature of loading or any approximation to the actual stress strain curve which everywhere has a lower stress for the same strain as the actual monotonic curve may be used. Either the shear or strain energy of distortion flow rule may be used. LP = Permissible load under stated emergency or fault condi-tions. LLB Lower bound limit load with yield point equal to 1.5 s,. = The " lower bound limit load" is here defined as that produced from the analysis of an ideally plastic (non-strain hardening) material where formations increase with no further increase in applied load. The lower bound load is one in which the materi-1 everywhere satis-fled equilibrium and nowhere exceeds the defined anterial yield - strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding to the uniaxial case. S, is the tabulated value of ASME III, or its equivalent, allowable stress at temperature. LPI Plastic instability load. The " plastic instability load" = is defined here as the load at which any load bearing sections begins to diminish its cross-sectional area at a faster rate than the strain hardening can accomodate the loss in area. This type analysis requires a true stress-true strain curve or a close approximation based on monotonic loading at the temperature of s loading, LUF Ultimate load from fracture analyzer. For components = r which involve sharp discontinuities (local theoretical stress concentration 3) the use of a " Fracture Mechanics" analysis where applicable utilizing measurements of plain strain fracture toughness may be applied to compute fracture loads. Correction for finite plastic zones and thickness effects 'as well as gross yielding may be necessary. The methods of linear elastic stress analysis may be used in the fracture analysis where its use is clearly conservative or supported by experimental evi-dence. Examples where " Fracture Mechanics" may be applied are for fillet welds or end of fatigue life crack population. LU* Ultimate load or loss of function load as determined from = or experiment. In usir.g this method, account shall be taken LLF of the dimensional tolerances which may exist between the x
T LGS TABLE C.4.4 (Continraad) PRIMARY STRESS LIMIT (Continued) . actual part and the tested part or parts as well as dif-ferences which may exist in the ultimate tensile strength of the actual part and the tested parts. The guide to be used in each of these areas is that the experimentally determined load shall use adjusted vaines to account for material properties and dimension variations, each of which has no greater probability than 0.1 of being exceeded in the actual part. DEFORMATION LIMIT CRIT 1PBT4** No. .Any One of (Only One Required) General Limit DI Permissible Deformation DP < 0.9 Analyzed Deformation % ~ SF,g, causing Loss of Function D2* Permissible Deformation DP < 1.0 Experimer:tal Deformation 57* ~ Sp Causing Loss of Function min DP Permissible Deformation under stated normal, upset, = emergency, or fault conditions. DLF3= Analyzed Deformation which would cause a system loss of function as delineated in the design specification. i DLFX= Experimentally Determined Inforanation which would cause a system _l.oss of fun _c_ti_on as delineated in the design i specification. " Loss of Function" can only be defined quite generally until attention is, focused on the component of interest. In cases of interest where deformation limits can affect the function of Class I equipment and components, they will be specifically dalineated. Examples where such limits apply are: ccantrol rod drive align-ment and clearances for proper insertior:, core support defor-mation causing fuel disarrangement, excess leakage of any com-ponsnt. BUCKLING STABILITY LIMIT * *
- No.
Any One of (Only One Required) General Limit B1 Pemissible Load LP < 2. 25 Code Normal Ever.t Permissible Load C# SFmin s t' A 11 v..*.. in 1671 Y
4 3 r LGS 1 TABLE C. 4. 4 (Continued) BUCKLING STABILITY LIMIT (Continued) No. Any One of (Oniv One Recuired) General Limit B2* Permissible Load LP < 1.0 5 ~ l Ultimate Buckling Collapse Load from LUB[SFg, Test l 33 Permissible Load l LP < 0.9 Stability Analysis Load % ~ SF,y, LP = Permissible Load under stated normal, upset, emergency, or fault conditions. Lar = Applicable code normal event permissible load. LUB Ultimate buckling collapse load,as determined from = experiment. In using this method, account shall be taken of the diagonal tolerances which may exist between the actual part and the tested part. The guide to be used in each of these areas is that the 4 experimentally determined load shall be adjusted to i account for material property and dimension variations, i each of which has no greater probability than 0.1 of being exceeded in the actual part. L5 = 3 Stability analysis load.. The ideal buckling analysis is of ten sensitive to otherwise minor deviations from ideal geometry and boundary conditions. These effects shall be accounted for in the analysis of the buckling stability loads. Examples of this are ovality in exter-nally pressurized shells or eccentricity of column members. FATIGUE LIMIT CRITERIA *** No. Any One of (Only One Reauired) Summation of fatigue damage usage with design and operation loads following Miner hypotheses **-shall not exceed F1 F2, or F3 as appropriate: Fl* Mean Fatigue cycle usage from analysis F 1 0.05 = 3 F2* Mean Fatigue cycle usage from test Fy 10.33 = F3 Design fatigue cycle usage from analysisi FD10 =
- Fatigue failure is defined here as the more limiting of:
.-~ ] ['__ ~ i LGS i TABLE C,4. 4 (Continued) FATIGUE LIMIT CRITERIA (Continued) 1. a 25 per. cent area reduction for a load carrying member ) which is required to function, or 2. excessive leakage 1 ) In the fatigue ev. luation the methods of linear electric stress analysis may be used when the 3S, range limit of ASME III has been met. If 3S, is not met, account will be taken of: (a) increases in local, strain concentration, (b) strain ratcheting, (c) redistribution of strain due to elastic-plastic effects. The February,1968, draft of the USAS B31.7 Piping Code may be used where appli-cable or detailed elastic-plastic methods may be used. With elastic-plastic methods, strain hardening may be used not to exceed in stress for the same strain, the j steady state cyclic strain hardening measured in a j smooth low cycle fatigue specimen at the average temp-erature of interest. ( l (
- Equations PS, P6, P7, D2, F1, P2, and B2 will not be applied unless supporting data are submitted for evaluation by the
) AEC staff.
- Minor, M.A., " Cumulative Damage in Fatigue," Journal of Applied Mechanics, Vol.12, Trans. ASME, Vol. 67, pp. A159-
) A164, Sept. 19 45. It is acceptable to use the ASME Section III Design Fatigue Curves in conjunction with a cumulative i usage factor of 1.0 (using Miner's hypothesis) in lieu of using the mean fatigue data curves with a limit on fatigue usage of 0. 05 tsince the two methods are approximately equiva-lont.
- Formal analysis required only where appropriate.
- Using method from Table C.4.5 C.4-13
'a'-
~~ 'c~ ~~ ~~ ~ ' ~ ~ t LGS m s( TABLE C.4.5 ,kj b CORE SUPPORT STRUCTURES, l' STRESS CATEGORIES AND LIMITS OF STRESS : FOR NORMAL AND UPSET CONDITIO1 PRIMARY STRESSES STRESS .M CAT E RY Membrane, P,(Notes 4,7,8) Bending P(Notes 4,7,8) Second 1 i) P, P *# a b P,ji' t Q S, 1.5S J l I Elastic Elastic Or Analysis Or Analysis l l Normal (Note 6) (tkete 6) ) -) .67 - 67 Upset Limit Limit or Analysis Or Analysis (Note 10) (Note 10) l .44L .44L ~ F T Test Test (Note 11) (Note 11) ( W 4
S 2 - Q3 [ 1.. w-g sr_+ par _ e c _,,. - s ~. - > -. : bw qnn~r-A'k~r:#=YWMDfN5(h $, rl. 6 t ' L*,' s
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S' I l C. 4' 5 t STRUCTURES ITS OF STRESS IENTENSITY f UPSET CONDITIONS k 1 l SECONDARY STRESSES PEAK STRESSES Membrane & Bending ,7,8) Secondary Q(Notes 2,4,6) Peak F (Notes 2,4,6) P,+P *O
- m*# *O*I b
b oD Elastic Elastic j o Or Analysis Analysis ) (Note 1) (Notes 3 s 9) i 8L I 1 i Limit s Or Analysis ' 0) (Note 5) e P,+P +Q+F b For Cycle Less Than 1000, Use Peak Elastic-Plastic 1) (Note 12) Fatf.gue (Notes 3,9,12) ) l C.4-14 July 30,1971 l
3 LGS TABLE C.4.5 NOTES NORMAL AND UPSET CONDITIONS NOTE 1 - This li2nitation applies to the range of stress in-tensity. When the secondar;* stress is due to a temperature excursion at the point at winich the stresses are being analyzed, the value of S. shall be taken as the. average of the S values ta5ulated in Tables N-421, N-422, and N-423 of ASME Boiler and Pressure Vessel Code, Section III, (ASME III) for the highest and the lowest temperature of the metal during the transient. When part of the secondary stress is due to mechanical load, the value of S, shall be taken as the S value for the h5ghest test-m perature of the metal during the transient. NOTE 2 - The stresses in Category Q are those parts of the total stress which are produced by thersaal gradia*=, structural discontinuities, etc., and do not include primary stresses which may also exist at the same point. It should be noted, however, that a detailed l stress analysis frequently gives the combination of primary and secondary stresses directly and, when appropriate, this calculated value represents the total of Pm+Pb + 0 and not o alone. similarly, if the stress in Category F is produc6d by a stress con-centration, the quantity F is the additional stress produced by the notch, over and above the nominal stress. For example, if a plate has a nneinal stress intensity, Pm = S, Pb = 0, O = 0 and a notch with a stress concentration K is introduced, then F = Pa (K - 1) and the peak stress intensity equals P, + P sa (K - 1) = KPm' NOTE 3 - S is obtained from the fatigue curves, Figures N-415 a of ASME III. The allowable stress intensity for the full range of fluctuation is 2 S - a NOTE 4 - The symbols P, P, Q, and F do not represent single m 3 quantities, but rather sets of six quantities repre-senting the six stress components a t 01' Ore Ttl' t Tir, and Trt* NOTE 5 - S denotes the structural action of shakedown load as t defined in par. N-412 (9) of ASME III calculated on a plastic basis as applied to a specific location on the structure. NOTE 6 - The triaxial stresses represent the algebraic sum of the three primary principal stresses (cy + Where uni-c2 + 03) for the combination of stress components. form tension loading is present triaxial stresses are limited to 4 S. m
- 9 O
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- Sb 9 Wh og g
y- - 7 7 ; --- --- - - - 7 LGS TABLE C.4.5 NOTES (Continued) NOTE 7 - For configurations where compressive stresses occur, the stress limits shall be revised to take into account critical buckling stresses (see par. N-410 ,l of ASME III). For. external pressure, the permissi-ble " equivalent stattic" external pressure shall be as specified by the rules of par. N-417.8 of ASME 'lI III. Where dynamic pressures are involved, the permissible external pressure shall be limited to 25% of the dynamic instability pressure. jj i, NOTE 8 - When loads are transiently applied, consideration f! should be given to the use of dynamic load amplifi-cation, and possible change in modulus of elasticity. j II NOTE 9 - In the fatigue data curves, where th'e number of ope-1 rating cycles are less than ten, use the S value j a for ten cycles; whe are greater than 10ge the number of operating cycles , use the s value for 106 a cycles. ,l WOTE 10 - LL is the lower bound limit load with yield point equal to 1.5 Sm (where S, is the tabulated value of l' allowable stress at temperature as contained in ASME l III). The " lower bound limit load" is here defined as that produced from the analysis of an ideally i plastic (non-strain hardening) material where deforun-ations increase with no further increase in applied load. The lower bound load is one in which the material everywhere satisfies equilibrium and nowhere exceeds the defined material yield strength using either a shear theory or a strain energy of distor-tion theory to relate multiaxial yielding to the uniaxial case. NOTE 11 - For normal and upset conditions, the limits on pri-mary membrane plus primary bending need not be satis-fled in a component if it can be shown from the test of a prototype or model that the specified loads i (dynamic or static equivalent) do not exceed 44% of L, where Lu is the ultimate load or the maximum load u or load combination used in the test. In using this method, account shall be taken of the size effect and dimensional tolerances which may exist between the actual part and the tested part or ~ parts as well as differences which may exist in the ultimate strength or other governing material properties of the actual part and the tested part to assure that the loads obtained from the test are a conservative represent-- ation of the load carrying capability of the actual coruponent under the postulated loading for Normal and Upset Conditions. l i C.4-16 .T o l.v 10. 1971
I E' LGS TABLE C.4.5 NOTES (Continued) NOTE 12 - The allowable value for the maximum range of this stress intensity is 3S except for cyclic events which occur less than,1000 times during the design life of the plant. For this exception, in lieu of meeting the 3S, limit, an elastic-plastic fatigue at alysis in accordance with ASME III or ASA B 31.7 may be performed to demonstrate that the cumulative fatigue usage attributable to the combination of these low cycle events plus all over cyclic events does not exceed a fatigue usage value of 1.0. 1 e
s LGS CORE SU PO UCTURES STRESS CATEGORIES AND LIMITS OF STRESS INTd FOR EMERGENCY CONDITIONS PRIMARY STRESSES STRESS SECONDARY S CATEGORY Membrane,Pa Bending, P3 Membraneil (Notes 1,2,610) (Notes 1,2810) Second P, P,+PB )-fi Elastic Elastic 9 1.5S, Analysis q .25 } Analysis [U (Note 3) (Note 3) J U Or Or 5 s O} f Limit Limit g Analysis 9 7 g Analysis 'ci G (Note 4) (Note 4) jj ti Emergency Or Or Evalua IN D' 8I Plastic Plastic Not R --1 1.5S Analysis .25S Analysis lj (Note 6) (Notes 5 [ & 6) )(.' Or Test H.6L, (Note 7) .5S (Note 5) u Or Or S)E Stress - l Ratio 6L* Tests-Analysis (Note 7) ~(Note 8) Or Stress-KS Ratio E Analysis g (Note 8) f
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i l LGS I TABLE C.4.'B SUPPORT STRUCTURES D LIMITS OF STRESS INTDISITY RGENCY CONDITIONS l SECONDARY STPPCCFC PEAK STRESSES ing, PB Maunbrane & Bending Peak 1,2&10) Secondary-Q F Elastic Analysis (Note 3) Limit l Analysis (Note 4) Evaluation Evaluation Plastic Analysis (tiotes 5 & 6) \\ (Note 5) u Tests (Note 7) S tre t.s-Ratio Analysis (Note 8) C.4-18 July 30,1971 t .M
,P LGS TABLE C.4.6 NOTES EMERGENCY CONDITIONS 1 i NOTE 1 - The symbols P, Pd, Q, and F do not represent single m quantities, but rather sets of six quantities repre-senting the six stress components o ' 8 t 1, are Tti Tir, and Trt' I NOTE 2 - For configurations where compressive stresses occur, the stress limits shall be revised to take into account critical buckling stress. For external pressure, the peratissible " equivalent static" exter-nal pressure shall be taken as 150 percent of that permitted by the rules of par. N-417.8 of ASME l Boiler and Pressure Vessel Code, Section III. Where dynamic pressures are involved, the permissible ex-ternal pressure shall satisfy the preceding require-ments or be limited to 50% of the dynamic instability pressure. NOTE 3 - The triaxial stresses represent +.be algebraic sum of the three primary principal stresses (al + c2 + 83) for the combination of stress components. Wriere uni-form tension loading is present, triaxial stresses should be limited to 6S,. NOTE 4 - L,iL the lower bound limit load with yield point y equal to 1.5 S (where S is the tabulated value of allowable stre.ss intensi,ty at temperature as contained in ASME III). The " lower bound limit load" is here defined as that produced from the analysis of an ideally plastic (non-strain hardening) material where deformations increase with no further increase in applied load. The lowet bound load is one in which the material everywhere satisfies equilibrium and no-where exceeds the defined material yield strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding to the uniaxial case. NOTE 5 - S is the ultimate strength at temperature. Multi-u axial effects on ultimate strength shall be considered. NG2E 6 - This plastic analysis uses an elastic plastic evalu-ated nominal primary stress. Strain hardening of the material may be used for the actual monotonic stress-strain curve which everywhere has a lower stress for the same strain as the actual monotonic curve may be used. Either the shear or strain energy of distortion flow rule shall be used to account for multiaxial effects. NOTE 7 - For emergency conditions, the stress limits need not be satisfied if it can be shown from the test of a proto-type or model that the specified loads (dynamic or C.4-19 July 30, 1971
/ LGS c = TABLE C.4.6 NOTES (Continued) stetic equivalent) do not exceed 60% of L is the ultimate load or the maximum load 8r, where L, load combination used in the test. In using this mehtod, l-account shall be taken of the sige effect and dimensional tolerances which may exist between the actual part and the tested part or partr as well as differences which may exist in the ultimate strength or other governing material properties of the actual part and the tested parts to assure that the loads obtained from the test are a conservative representation of the load carryi:29 j capability of the actual component under postulated ~ loading for emergency conditions. NOTE 8 - Stress ratio is a method of plastic analysis which uses the stress ratio combinations (combination of stress that consider the ratio of the actual stress to the allowable plastic or clastic stress) to comi-pute the maximum
- load a strain hardening material can carry.
K is defined as the Section Factor; S 28 for pr ry merane loading. E M NOTE 9 - Where deformation is of concern in a component, the deformation shall be limited to two-thirds the value given for Emergency conditions in the Design Specifi-cation. NOTE 10 - When loads are transiently applied, consideration should be given to the use of dynamic load amplifica-tion and possible change in modulus of elasticity. \\ I l q C.4-20 July 30/1971 i 1
.e LGS T_ABLE C. 4. 7 CORE SUPPORT STRUCTUI-STRESS CATEGORIES AND LIMITS OF S')i i FOR FAULTED CONDITION l PRIMARY STRESSES STRESS Membrane P, (Notes 1,263) Bending Pb (Notes P, P, + PB i Elastic Elal Analysis 9 Anai Or or Faulted .758u I (Note 5) j .3 { (Note 9) Or Or Limit
- Plar,
.3 i Analysis .75 u l Ana]l 8 (Note 4) (Not Or or Plastic ,' 67S } Analysis .81, I Test (Noti u (Notes S&6 Or Or Stre gT:: gg xs 2) F Or (Nota Stress-Ratio SF Analysis (Note 8) I 'I
wS s LE C.4.7 7 j RES LIMITS OF STRESS INTENSITY .gg TED CONDITIONS ,g 1 -) SECONDARY STRESSES PEAK STRESSES i M rane & W ing Peak ig Pb (Notes 1,2,&3) 1,2 Secondary-Q F Py Elastic 'act j 3.0Sa l gg aly G .3 Evaluation Evaluation Not Required Not Required
- 1**Di
'8t 8Ql .75 u Analysis (Notes 5&6) C r Tests \\ to I (Note 7) Stress- 'UE i KS Ratio fo F U/U Analysis (Note 8) to i e C.4-21 July 30,1971
- , a LGS
\\ TABLE C.4.7 NOTES FAULTED CONDITIONS l' NOTE 1 - The symbols Pm, Pb, Q, and F do not represent quantitles i but rather sets of six quantities representing the six stress components, Ot, oi, o, Tti, Tir, and Trt. r i NOTE 2 - When loads are transiently applied, consideration should be given to the use of dynamic load amplification and possible changes in modulus of elasticity. J l NOTE 3 - For configurations where compressive stresses occur, the stress limits shall be revised to take into account cri - I tical buckling stresses. For external pressure, the permissible " equivalent static" external pressure shall be taken as 2.5 times that given by the rules of par. I l N-417.8 of ASME Boiler and Pressure Vessel Code Section III. Where dynamic pressur..s are involved, the permiss-f ible external pressure shall satisfy the preceding } j requirements or shall be limite<1 to 75% of the dynamic ( instability pressure. I NOTE 4 - LL is the lower bound limit load with yield point equal to 1.5 Sm (where Sm is the tabulated value of allowable stress intensity at temperature as contained in ASME III). The " lower bound limit load" is here defined as that produced from the analysis of an ideally plastic (non-strain hardening) material where deformations increase with no further increase in applied load. The lower bound load is one in which the material everywhere satis-fies equilibrium and nowhere exceeds the defined material l yield strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding ) [ to the uniaxial case. NOTE 5 - Su is the ultimate strength at temperature. Multiaxial ,1 effects on ultimate strength shall be considered, NOTE 6 - This plastic analysis uses an elastic-plastic evaluated i nominal primary stress. Strain hardening of the mater al may be used for the actual monotonic stress-strain curve at the temperature of loading or any approximation to the actual stress-strain curve which everywhere has a lower stress for the same strain as the actual monotonic curve may be used. Either the maximum shear stress or strain energy of distortion flow rule shall be used to account for multiaxial effects. C.4-22 July 30, 1971
I 7 LGS TABLE C. 4. 7 NO'"ES (Continued) i NOTE 7 - For Faulted Conditions, the stress limits need not be satisfied if it can be shown from the test of a prototype or model that the specified loads (dynamic or static equivalent) do not exceed 80% of Ly, where Ly is the ultimate load or load combination used in the test. l In using this method, account shall be taken of the size effect and dimensional tolerances as well as differences which may exist in the ultimate strength or other govern-ing material properties of the actual part and the tested parts to assure that the loads obtained from the test are a conservative representation of the load carrying capa-bility of the actual component under postulated loading for Faulted Condition. l NOTE 8 - Stress ratio is method of plastic analysis which uses the stress ratio combinations (combination of stresses that l consider the ratio of the actual stress to the allowable plastic or elastic stress) to compute the maximum load a { strain hardenint material can carry. K is defined as the Section Factor; Sy is the lesser of 2.4 S, c,e 0.75 S fD# primary membrane loading. u NOTE 9 - Where deformation is of concern in a component, the defor-mation shall' be limited to 80% of the value given for Faulted Conditions in the Design Specifications. \\ I C.4-23 Ju'$y 30,1971 i
i LGS
- I QUESTION 5.20 State the design differential pressure across the floor at ele-vation 236'00" (Figure 5.2.1) under loss-of-coolant accident conditions.
Also, indicate what initial and subsequent testing of the floor with regard to strength and leakage will be per-formed, and how it will be carried out. Specify the maximum allowables leakage which will not result in overpressure of either upper or lower compartments and state the allowable design leakage through the floor. Include the method and assumptions by which these leakage values are determined. ANSWER: The design pressure conditions of 55 psig in the drywell and 25 psig in the suppression chamber have been selected based upon l design basis loss-of-coolant accident. "hese conditions define a 30 psi design differential pressure across the drywell floor slab which envelopes the most severe loading of this component. l Structural and pressure integrity tests are to be performed j prior to plant operation and additional pressure integrity tests may be made subsequently during plant shutdown. Initial tests are to be conducted at 115% of the following design con-ditions: a. A design presrure condition of 55 psig in both the drywell and suppression chamber. b. A design pressure condition of 55 psig in the drywell and 25 psig in the suppression chamber. The differential pressure test of the drywell floor slab de-scribed in item (b) above is to be accomplished by capping the downcomers above the drywell floor slab upper surface. A liner plate, as described in the answer to Question 5.21, has been added to the upper surface of the drywell floor slab. This liner plate is of the same material and meets the same quality assurance requirements as the liner plate at the pri-mary containment boundary. Details of the drywell floor slab liner plate are shown in PSAR Supplement 3, Figure 05.20.1, dated June 1971. Based upon the information supplied in re-sponse to Question 14.12 submitted in PSAR Supplement No. 5, it can be concluded that only large drywell floor slab bypass (break) areas (exceeding about 2 square feet in area) will re-sult in overpressurization of the containment. Bypass areas approaching this magnitude will be detected during the struc-tural and pressure integrity test described above. Therefore, no quantitative leakage tests across the drywell floor slab will be performed. 10 July 30, 1971 ei
LGS However periodic low differential pressure (less than 4 paid) tests will be conducted subsequent to initial startup to insure no gross leakage path exists between the drywell and suppression chamber gas space. I 6 -l i 10a I July 30, 1971 k .i}}